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Sample records for vessel burst test

  1. Phenomenological vessel burst investigations

    International Nuclear Information System (INIS)

    Hippelein, K.W.; Julisch, P.; Muz, J.; Schiedermaier, J.

    1985-07-01

    Fourteen burst experiments have been carried out using vessels with circumferential and longitudinal flaws, for investigation of the fracture behaviour, i.e. the time-related fracture opening. The vessels had dimensions (outer diameter x wall thickness = 800 x 47 mm) which correspond to the dimensions of the main coolant piping of a 1300 MW e PWR. The test specimens had been made of the base-safe material 20 MnMoNi 55 and of a special, 22 NiMoCr 37 base alloy. The experimental conditions with regard to pressure and temperature have been chosen so as to correspond to normal operating conditions of a PWR (p∝17.5 MPa, T∝300 0 C), i.e. the flaws have been so dimensioned that failure was to be expected at a pressure of p∝17.5 MPa. As a rule, water has been used as the pressure medium, or in some cases air, in order to influence the time-dependent pressure decrease. Fluid and structural dynamics calculations have also been made. In order to determine the impact of a fast propagating crack on the leak-to-fracture curve, which normally is defined by quasistationary experiments, suitable tests have been made with large-volume, cylindrical vessels (outer diameter x wall thickness x length = 3000 x 21 x 14000 mm) made of the material WSt E 43. The leak-before-fracture criterion has been confirmed. (orig./HP) [de

  2. Bursting tests on pressure vessels with cracks differing in configuration and location

    International Nuclear Information System (INIS)

    Stahlberg, R.

    1978-01-01

    For assessing the safety of nuclear pressure vessels exhibiting cracks, bursting test were carried out on a series of medium-size pressure vessels with and without welded nozzles and exhibiting cracks differing in configuration and location. The linear-elastic approach proved to be sufficiently accurate for straight strain conditions up to the onset of general yielding. Other analytical methods were successfully used to cover the plastic region. (orig.) [de

  3. Device for the burst protection of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Daublebsky, P.

    1976-01-01

    The burst protection device has a hood over top and bottom of the pressure vessel with superimposed hinged supports lying in their turn against supporting rings which are connected with each other by vertical bracing. It is proposed to place an intermediate layer between hoods and vertical bracing absorbing thermal stresses, i.e. deforming plastically with gradually increasing pressure, but behaving like a rigid body in the case of shock loads. As a material lead e.g. is proposed. (UWI) [de

  4. Burst Test Qualification Analysis of DWPF Canister-Plug Weld

    International Nuclear Information System (INIS)

    Gupta, N.K.; Gong, Chung.

    1995-02-01

    The DWPF canister closure system uses resistance welding for sealing the canister nozzle and plug to ensure leak tightness. The welding group at SRTC is using the burst test to qualify this seal weld in lieu of the shear test in ASME B ampersand PV Code, Section IX, paragraph QW-196. The burst test is considered simpler and more appropriate than the shear test for this application. Although the geometry, loading and boundary conditions are quite different in the two tests, structural analyses show similarity in the failure mode of the shear test in paragraph QW-196 and the burst test on the DWPF canister nozzle Non-linear structural analyses are performed using finite element techniques to study the failure mode of the two tests. Actual test geometry and realistic stress strain data for the 304L stainless steel and the weld material are used in the analyses. The finite element models are loaded until failure strains are reached. The failure modes in both tests are shear at the failure points. Based on these observations, it is concluded that the use of a burst test in lieu of the shear test for qualifying the canister-plug weld is acceptable. The burst test analysis for the canister-plug also yields the burst pressures which compare favorably with the actual pressure found during burst tests. Thus, the analysis also provides an estimate of the safety margins in the design of these vessels

  5. Prestressed cast iron pressure vessels as burst-proof pressure vessels for innovative nuclear applications

    International Nuclear Information System (INIS)

    Froehling, W.; Boettcher, A.; Bounin, D.; Steinwarz, W.; Geiss, M.; Trauth, M.

    2000-01-01

    The amendment to the German Atomic Energy Act from July 28, 1994 requires that events 'whose occurrence is practically excluded by the measures against damages', i.e. events of the category residual risk, must not necessitate far reaching protective measures outside the plant. For a conventional reactor pressure vessel, the residual risk consists in the very small probability of a catastrophic failure (formation of a large fracture opening, bursting of the vessel). With a prestressed cast iron vessel (PCIV), the formation of a large fracture opening or bursting of the vessel, respectively, is impossible due to its design properties. Against this background the possibility of the use of this type of pressure vessel for lightwater reactors has been studied in the frame of a 'Working Group for Innovative Nuclear Technology', founded by different research institutes and industrial companies. Furthermore, it has been studied whether the use of the PCIV support the realization of a corecatcher system. The results are presented in this report. Already many years earlier, Siempelkamp has performed industrial development and Forschungszentrum Juelich related experimental and theoretical safety research for the PCIV as an innovative, bust-proof pressure vessel concept. This development of the PCIV as well as its safety properties are also presented in a conclusive manner. (orig.) [de

  6. Acrylic vessel cleaning tests

    International Nuclear Information System (INIS)

    Earle, D.; Hahn, R.L.; Boger, J.; Bonvin, E.

    1997-01-01

    The acrylic vessel as constructed is dirty. The dirt includes blue tape, Al tape, grease pencil, gemak, the glue or residue form these tapes, finger prints and dust of an unknown composition but probably mostly acrylic dust. This dirt has to be removed and once removed, the vessel has to be kept clean or at least to be easily cleanable at some future stage when access becomes much more difficult. The authors report on the results of a series of tests designed: (a) to prepare typical dirty samples of acrylic; (b) to remove dirt stuck to the acrylic surface; and (c) to measure the optical quality and Th concentration after cleaning. Specifications of the vessel call for very low levels of Th which could come from tape residues, the grease pencil, or other sources of dirt. This report does not address the concerns of how to keep the vessel clean after an initial cleaning and during the removal of the scaffolding. Alconox is recommended as the cleaner of choice. This acrylic vessel will be used in the Sudbury Neutrino Observatory

  7. Test of 6-in.-thick pressure vessels. Series 3: intermediate test vessel V-7

    International Nuclear Information System (INIS)

    Merkle, J.G.; Robinson, G.C.; Holz, P.P.; Smith, J.E.; Bryan, R.H.

    1976-08-01

    The test of intermediate test vessel V-7 was a crack-initiation fracture test of a 152-mm-thick (6-in.), 990-mm-OD (39-in.) vessel of ASTM A533, grade B, class 1 steel plate with a sharp outside surface flaw 457 mm (18 in.) long and about 135 mm (5.3 in.) deep. The vessel was heated to 91 0 C (196 0 F) and pressurized hydraulically until leakage through the flaw terminated the test at a peak pressure of 147 MPa (21,350 psi). Fracture toughness data obtained by testing precracked Charpy-V and compact-tension specimens machined from a prolongation of the cylindrical test shell were used in pretest analyses of the flawed vessel. The vessel, as expected, did not burst. Upon depressurization, the ruptured ligament closed so as to maintain static pressure without leakage at about 129 MPa

  8. Burst pressure investigation of filament wound type IV composite pressure vessel

    Science.gov (United States)

    Farhood, Naseer H.; Karuppanan, Saravanan; Ya, H. H.; Baharom, Mohamad Ariff

    2017-12-01

    Currently, composite pressure vessels (PVs) are employed in many industries such as aerospace, transportations, medical etc. Basically, the use of PVs in automotive application as a compressed natural gas (CNG) storage cylinder has been growing rapidly. Burst failure due to the laminate failure is the most critical failure mechanism for composite pressure vessels. It is predominantly caused by excessive internal pressure due to an overfilling or an overheating. In order to reduce fabrication difficulties and increase the structural efficiency, researches and studies are conducted continuously towards the proper selection of vessel design parameters. Hence, this paper is focused on the prediction of first ply failure pressure for such vessels utilizing finite element simulation based on Tsai-Wu and maximum stress failure criterions. The effects of laminate stacking sequence and orientation angle on the burst pressure were investigated in this work for a constant layered thickness PV. Two types of winding design, A [90°2/∓θ16/90°2] and B [90°2/∓θ]ns with different orientations of helical winding reinforcement were analyzed for carbon/epoxy composite material. It was found that laminate A sustained a maximum burst pressure of 55 MPa for a sequence of [90°2/∓15°16/90°2] while the laminate B returned a maximum burst pressure of 45 MPa corresponding to a stacking sequence of [90°2/±15°/90°2/±15°/90°2/±15° ....] up to 20 layers for a constant vessel thickness. For verification, a comparison was done with the literature under similar conditions of analysis and good agreement was achieved with a maximum difference of 4% and 10% for symmetrical and unsymmetrical layout, respectively.

  9. Model tests for prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Stoever, R.

    1975-01-01

    Investigations with models of reactor pressure vessels are used to check results of three dimensional calculation methods and to predict the behaviour of the prototype. Model tests with 1:50 elastic pressure vessel models and with a 1:5 prestressed concrete pressure vessel are described and experimental results are presented. (orig.) [de

  10. Burst protection for reactor pressure vessels using a hinged support bearing

    International Nuclear Information System (INIS)

    Michel, E.; Maritsch, F.

    1976-01-01

    The invention deals with a simplification of the design and manufacture and the way of controlling a hinged support bearing used as burst protection. The pure pressure load of the, e.g., 32 hinged supports distributed along the circumference of the pressure vessel head is achieved in the braced state with little control effort by a pure rotating motion caused pneumatically or hydraulically. The hinged supports are inclined by about 45 0 upwards/outwards in the braced state and with their cap-shaped head and foot are selflocking by pivoted between a supporting structure, firmly connected with the building, and a fishing ring. (TK) [de

  11. Test of 6-in.-thick pressure vessels. Series 3: intermediate test vessel V-7A under sustained loading

    International Nuclear Information System (INIS)

    Bryan, R.H.; Cate, T.M.; Holz, P.P.; King, T.A.; Merkle, J.G.; Robinson, G.C.; Smith, G.C.; Smith, J.E.; Whitman, G.D.

    1978-01-01

    HSST intermediate test vessel V-7 was repaired after being tested hydrostatically to leakage and was retested pneumatically as vessel V-7A. Except for the method of applying the load, the conditions in both tests were nearly identical. In each case, a sharp outside surface flaw 547 mm long (18 in.) by about 135 mm deep (5.3 in.) was prepared in the 152-mm-thick (6-in.) test cylinder of A533, grade B, class 1 steel. The inside surface of vessel V-7A was sealed in the region of the flaw by a thin metal patch so that pressure could be sustained after rupture. Vessel V-7A failed by rupture of the flaw ligament without burst, as expected. Rupture occurred at 144.3 MPa (20.92 ksi), after which pressure was sustained for 30 min without any indication of instability. The rupture pressure of vessel V-7A was about 2 percent less than that of vessel V-7

  12. Creep Burst Testing of a Woven Inflatable Module

    Science.gov (United States)

    Selig, Molly M.; Valle, Gerard D.; James, George H.; Oliveras, Ovidio M.; Jones, Thomas C.; Doggett, William R.

    2015-01-01

    A woven Vectran inflatable module 88 inches in diameter and 10 feet long was tested at the NASA Johnson Space Center until failure from creep. The module was pressurized pneumatically to an internal pressure of 145 psig, and was held at pressure until burst. The external environment remained at standard atmospheric temperature and pressure. The module burst occurred after 49 minutes at the target pressure. The test article pressure and temperature were monitored, and video footage of the burst was captured at 60 FPS. Photogrammetry was used to obtain strain measurements of some of the webbing. Accelerometers on the test article measured the dynamic response. This paper discusses the test article, test setup, predictions, observations, photogrammetry technique and strain results, structural dynamics methods and quick-look results, and a comparison of the module level creep behavior to the strap level creep behavior.

  13. Proof testing of an explosion containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Esparza, E.D. [Esparza (Edward D.), San Antonio, TX (United States); Stacy, H.; Wackerle, J. [Los Alamos National Lab., NM (United States)

    1996-10-01

    A steel containment vessel was fabricated and proof tested for use by the Los Alamos National Laboratory at their M-9 facility. The HY-100 steel vessel was designed to provide total containment for high explosives tests up to 22 lb (10 kg) of TNT equivalent. The vessel was fabricated from an 11.5-ft diameter cylindrical shell, 1.5 in thick, and 2:1 elliptical ends, 2 in thick. Prior to delivery and acceptance, three types of tests were required for proof testing the vessel: a hydrostatic pressure test, air leak tests, and two full design charge explosion tests. The hydrostatic pressure test provided an initial static check on the capacity of the vessel and functioning of the strain instrumentation. The pneumatic air leak tests were performed before, in between, and after the explosion tests. After three smaller preliminary charge tests, the full design charge weight explosion tests demonstrated that no yielding occurred in the vessel at its rated capacity. The blast pressures generated by the explosions and the dynamic response of the vessel were measured and recorded with 33 strain channels, 4 blast pressure channels, 2 gas pressure channels, and 3 displacement channels. This paper presents an overview of the test program, a short summary of the methodology used to predict the design blast loads, a brief description of the transducer locations and measurement systems, some of the hydrostatic test strain and stress results, examples of the explosion pressure and dynamic strain data, and some comparisons of the measured data with the design loads and stresses on the vessel.

  14. Multirods burst tests under loss-of-coolant conditions

    International Nuclear Information System (INIS)

    Kawasaki, S.; Uetsuka, H.; Furuta, T.

    1983-01-01

    In order to know the upper limit of coolant flow area restriction in a fuel assembly under loss-of-coolant accidents in LWRs, burst tests of fuel bundles were performed. Each bundle consisted of 49 rods(7x7 rods), and bursts were conducted in flowing steam. In some cases, 4 rods were replaced by control rods with guide tubes in a bundle. After the burst, the ballooning behavior of each rod and the degree of coolant flow area restriction in the bundle were measured. Ballooning behavior of rods and degree of coolant flow channel restriction in bundles with control rods were not different from those without control rods. The upper limit of coolant flow channel restriction under loss-of-coolant conditions was estimated to be about 80%. (author)

  15. On the hydrostatic test for nuclear vessels

    International Nuclear Information System (INIS)

    Palmero, A.

    1979-01-01

    A comparison of the pressure test requirements, namely specified values of pressure and temperature, for nuclear vessels designed and constructed according to the ASME Code and Spanish Rules is presented. Also the relationship of the design criteria and the pressure test requirements is indicated with a particular emphasis on the test temperature in order to avoid brittle behaviour of the materials. (author)

  16. Results of steel containment vessel model test

    International Nuclear Information System (INIS)

    Luk, V.K.; Ludwigsen, J.S.; Hessheimer, M.F.; Komine, Kuniaki; Matsumoto, Tomoyuki; Costello, J.F.

    1998-05-01

    A series of static overpressurization tests of scale models of nuclear containment structures is being conducted by Sandia National Laboratories for the Nuclear Power Engineering Corporation of Japan and the US Nuclear Regulatory Commission. Two tests are being conducted: (1) a test of a model of a steel containment vessel (SCV) and (2) a test of a model of a prestressed concrete containment vessel (PCCV). This paper summarizes the conduct of the high pressure pneumatic test of the SCV model and the results of that test. Results of this test are summarized and are compared with pretest predictions performed by the sponsoring organizations and others who participated in a blind pretest prediction effort. Questions raised by this comparison are identified and plans for posttest analysis are discussed

  17. Testing the performance of a blind burst statistic

    Energy Technology Data Exchange (ETDEWEB)

    Vicere, A [Istituto di Fisica, Universita di Urbino (Italy); Calamai, G [Istituto Nazionale di Fisica Nucleare, Sez. Firenze/Urbino (Italy); Campagna, E [Istituto Nazionale di Fisica Nucleare, Sez. Firenze/Urbino (Italy); Conforto, G [Istituto di Fisica, Universita di Urbino (Italy); Cuoco, E [Istituto Nazionale di Fisica Nucleare, Sez. Firenze/Urbino (Italy); Dominici, P [Istituto di Fisica, Universita di Urbino (Italy); Fiori, I [Istituto di Fisica, Universita di Urbino (Italy); Guidi, G M [Istituto di Fisica, Universita di Urbino (Italy); Losurdo, G [Istituto Nazionale di Fisica Nucleare, Sez. Firenze/Urbino (Italy); Martelli, F [Istituto di Fisica, Universita di Urbino (Italy); Mazzoni, M [Istituto Nazionale di Fisica Nucleare, Sez. Firenze/Urbino (Italy); Perniola, B [Istituto di Fisica, Universita di Urbino (Italy); Stanga, R [Istituto Nazionale di Fisica Nucleare, Sez. Firenze/Urbino (Italy); Vetrano, F [Istituto di Fisica, Universita di Urbino (Italy)

    2003-09-07

    In this work, we estimate the performance of a method for the detection of burst events in the data produced by interferometric gravitational wave detectors. We compute the receiver operating characteristics in the specific case of a simulated noise having the spectral density expected for Virgo, using test signals taken from a library of possible waveforms emitted during the collapse of the core of type II supernovae.

  18. Testing of a steel containment vessel model

    International Nuclear Information System (INIS)

    Luk, V.K.; Hessheimer, M.F.; Matsumoto, T.; Komine, K.; Costello, J.F.

    1997-01-01

    A mixed-scale containment vessel model, with 1:10 in containment geometry and 1:4 in shell thickness, was fabricated to represent an improved, boiling water reactor (BWR) Mark II containment vessel. A contact structure, installed over the model and separated at a nominally uniform distance from it, provided a simplified representation of a reactor shield building in the actual plant. This paper describes the pretest preparations and the conduct of the high pressure test of the model performed on December 11-12, 1996. 4 refs., 2 figs

  19. Eddy current testing of composite pressure vessels

    Science.gov (United States)

    Casperson, R.; Pohl, R.; Munzke, D.; Becker, B.; Pelkner, M.

    2018-04-01

    The use of composite pressure vessels instead of conventional vessels made of steel or aluminum grew strongly over the last decade. The reason for this trend is the tremendous weight saving in the case of composite vessels. However, the long-time behavior is not fully understood for filling and discharging cycles and creep strength and their influence on the CFRP coating (carbon fiber reinforced plastics) and the internal liner (steel, aluminum, or plastics). The CFRP ensures the pressure resistance while the inner liner is used as a container for liquid or gas. To overcome the missing knowledge of aging, BAM started an internal project to investigate degradation of these material systems. Therefore, applicable testing methods like eddy current testing are needed. Normally, high-frequency eddy current testing (HF-ET, f > 10 MHz) is deployed for CFRP due to its low conductivity of the fiber, which is in the order of 0.01 MS/s, and the capacitive coupling between the fibers. Nevertheless, in some cases conventional ET can be applied. We show a concise summary of studies on the application of conventional ET of composite pressure vessels.

  20. Testing Einstein's Equivalence Principle With Fast Radio Bursts.

    Science.gov (United States)

    Wei, Jun-Jie; Gao, He; Wu, Xue-Feng; Mészáros, Peter

    2015-12-31

    The accuracy of Einstein's equivalence principle (EEP) can be tested with the observed time delays between correlated particles or photons that are emitted from astronomical sources. Assuming as a lower limit that the time delays are caused mainly by the gravitational potential of the Milky Way, we prove that fast radio bursts (FRBs) of cosmological origin can be used to constrain the EEP with high accuracy. Taking FRB 110220 and two possible FRB/gamma-ray burst (GRB) association systems (FRB/GRB 101011A and FRB/GRB 100704A) as examples, we obtain a strict upper limit on the differences of the parametrized post-Newtonian parameter γ values as low as [γ(1.23  GHz)-γ(1.45  GHz)]radio energies, improving by 1 to 2 orders of magnitude the previous results at other energies based on supernova 1987A and GRBs.

  1. Applicability of modified burst test data to reactivity initiated accident

    Energy Technology Data Exchange (ETDEWEB)

    Yueh, K., E-mail: yuehky@hotmail.com

    2017-05-15

    A comprehensive irradiated cladding mechanical property dataset was generated by a recently developed modified burst test (MBT) under reactivity initiated accident (RIA) loading conditions [1,2]. The test data contains a wide range of test conditions that could bridge the gap between fast transient test reactor data (short pulse and/or low temperature) and prototypical commercial reactor conditions. This paper documents an evaluation performed to demonstrate the applicability of the MBT data to fuel cladding performance under RIA conditions. The current effort includes a comparison of calculated fuel cladding failure/burst strain for tests conducted at the Japan Atomic Energy Agency's (JAEA) Nuclear Safety Research Reactor (NSRR) to the MBT dataset, and an evaluation of potential mechanisms on how some NSRR tests survived beyond the cladding loading capacity. A simple shell model, coupled with temperature output from the Falcon fuel performance code, was used to calculate the fuel pellet thermal expansion of NSRR tests at the point of failure. The calculated fuel pellet thermal expansion correlates well directly with the MBT data at similar loading conditions. A 3-dimensional (3D) finite element analysis (FEA) model was used to evaluate fuel movement potential during a RIA. The evaluation indicates fuel relocation into the pellet chamfer and later into the dish is possible once a temperature threshold is reached before cladding failure and thus could significantly increase the fuel rod energy absorption capacity in a RIA event.

  2. A Static Burst Test for Composite Flywheel Rotors

    Science.gov (United States)

    Hartl, Stefan; Schulz, Alexander; Sima, Harald; Koch, Thomas; Kaltenbacher, Manfred

    2016-06-01

    High efficient and safe flywheels are an interesting technology for decentralized energy storage. To ensure all safety aspects, a static test method for a controlled initiation of a burst event for composite flywheel rotors is presented with nearly the same stress distribution as in the dynamic case, rotating with maximum speed. In addition to failure prediction using different maximum stress criteria and a safety factor, a set of tensile and compressive tests is carried out to identify the parameters of the used carbon fiber reinforced plastics (CFRP) material. The static finite element (FE) simulation results of the flywheel static burst test (FSBT) compare well to the quasistatic FE-simulation results of the flywheel rotor using inertia loads. Furthermore, it is demonstrated that the presented method is a very good controllable and observable possibility to test a high speed flywheel energy storage system (FESS) rotor in a static way. Thereby, a much more expensive and dangerous dynamic spin up test with possible uncertainties can be substituted.

  3. Power Burst Facility Severe Fuel Damage test series

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  4. Biaxial Loading Tests for steel containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Miyagawa, T. [Nuclear Power Engineering Corp., Tokyo (Japan); Wright, D.J.; Arai, S.

    1999-07-01

    The Nuclear Power Engineering Corporation (NUPEC) has conducted a 1/10 scale of the steel containment vessel (SCV) test for the understanding of ultimate structural behavior beyond the design pressure condition. Biaxial Loading Tests were supporting tests for the 1/10 scale SCV model to evaluate the method of estimating failure conditions of thin steel plates under biaxial loading conditions. The tentative material models of SGV480 and SPV490 were obtained. And the behavior of SGV480 and SPV490 thin steel plates under biaxial loading conditions could be well simulated by FE-Analyses with the tentative material models and Mises constitutive law. This paper describes the results and the evaluations of these tests. (author)

  5. Biaxial Loading Tests for steel containment vessel

    International Nuclear Information System (INIS)

    Miyagawa, T.; Wright, D.J.; Arai, S.

    1999-01-01

    The Nuclear Power Engineering Corporation (NUPEC) has conducted a 1/10 scale of the steel containment vessel (SCV) test for the understanding of ultimate structural behavior beyond the design pressure condition. Biaxial Loading Tests were supporting tests for the 1/10 scale SCV model to evaluate the method of estimating failure conditions of thin steel plates under biaxial loading conditions. The tentative material models of SGV480 and SPV490 were obtained. And the behavior of SGV480 and SPV490 thin steel plates under biaxial loading conditions could be well simulated by FE-Analyses with the tentative material models and Mises constitutive law. This paper describes the results and the evaluations of these tests. (author)

  6. Testing Einstein's Equivalence Principle With Fast Radio Bursts

    Science.gov (United States)

    Wei, Jun-Jie; Gao, He; Wu, Xue-Feng; Mészáros, Peter

    2015-12-01

    The accuracy of Einstein's equivalence principle (EEP) can be tested with the observed time delays between correlated particles or photons that are emitted from astronomical sources. Assuming as a lower limit that the time delays are caused mainly by the gravitational potential of the Milky Way, we prove that fast radio bursts (FRBs) of cosmological origin can be used to constrain the EEP with high accuracy. Taking FRB 110220 and two possible FRB/gamma-ray burst (GRB) association systems (FRB/GRB 101011A and FRB/GRB 100704A) as examples, we obtain a strict upper limit on the differences of the parametrized post-Newtonian parameter γ values as low as [γ (1.23 GHz )-γ (1.45 GHz )] <4.36 ×10-9. This provides the most stringent limit up to date on the EEP through the relative differential variations of the γ parameter at radio energies, improving by 1 to 2 orders of magnitude the previous results at other energies based on supernova 1987A and GRBs.

  7. EDS V25 containment vessel explosive qualification test report.

    Energy Technology Data Exchange (ETDEWEB)

    Rudolphi, John Joseph

    2012-04-01

    The V25 containment vessel was procured by the Project Manager, Non-Stockpile Chemical Materiel (PMNSCM) as a replacement vessel for use on the P2 Explosive Destruction Systems. It is the first EDS vessel to be fabricated under Code Case 2564 of the ASME Boiler and Pressure Vessel Code, which provides rules for the design of impulsively loaded vessels. The explosive rating for the vessel based on the Code Case is nine (9) pounds TNT-equivalent for up to 637 detonations. This limit is an increase from the 4.8 pounds TNT-equivalency rating for previous vessels. This report describes the explosive qualification tests that were performed in the vessel as part of the process for qualifying the vessel for explosive use. The tests consisted of a 11.25 pound TNT equivalent bare charge detonation followed by a 9 pound TNT equivalent detonation.

  8. Testing and Performance of UFFO Burst Alert & Trigger Telescope

    DEFF Research Database (Denmark)

    Rípa, Jakub; Bin Kim, Min; Lee, Jik

    2014-01-01

    The Ultra-Fast Flash Observatory pathfinder (UFFO-p) is a new space mission dedicated to detect Gamma-Ray Bursts (GRBs) and rapidly follow their afterglows in order to provide early optical/ultraviolet measurements. A GRB location is determined in a few seconds by the UFFO Burst Alert & Trigger t...

  9. Elastic-plastic failure analysis of pressure burst tests of thin toroidal shells

    International Nuclear Information System (INIS)

    Jones, D.P.; Holliday, J.E.; Larson, L.D.

    1998-07-01

    This paper provides a comparison between test and analysis results for bursting of thin toroidal shells. Testing was done by pressurizing two toroidal shells until failure by bursting. An analytical criterion for bursting is developed based on good agreement between structural instability predicted by large strain-large displacement elastic-plastic finite element analysis and observed burst pressure obtained from test. The failures were characterized by loss of local stability of the membrane section of the shells consistent with the predictions from the finite element analysis. Good agreement between measured and predicted burst pressure suggests that incipient structural instability as calculated by an elastic-plastic finite element analysis is a reasonable way to calculate the bursting pressure of thin membrane structures

  10. Results of gap conductance tests in the power burst facility

    International Nuclear Information System (INIS)

    Garner, R.W.; Sparks, D.T.

    1977-01-01

    Light water reactor (LWR) fuel rod behavior studies are being conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc. These studies are being performed under contract to the Energy Research and Development Adminstration at the Idaho National Engineering Laboratory (INEL), as part of the Nuclear Regulatory Commission's Water Reactor Safety Research Fuel Behavior Program. Experimental data for verification of analytical models developed to predict light water nuclear fuel rod behavior under normal and postulated accident conditions are being obtained from a variety of in-reactor and out-of-reactor experiments. This paper summarizes the results of tests performed in the Power Burst Facility (PBF) to obtain data from which the thermal response, gap conductance, and stored energy of LWR fuel rods can be determined. Primary objectives of the PBF gap conductance test program are (a) to obtain data on a variety of pressurized water reactor (PWR) and boiling water reactor (BWR) fuel rod designs, under a wide range of operating conditions, from which gap conductance values can be determined and (b) to evaluate experimentally the power oscillation method for measuring the gap conductance and thermal response of a fresh or burned LWR fuel rod. Tests have been performed with both irradiated and unirradiated PWR-type fuel and with fresh BWR-type fuel rods. Some PWR rod test results are described, and the thermal response data from BWR rod tests are discussed in greater detail. Comparisons are made of gap conductance values determined by the tests with analytically calculated values using the Fuel Rod Analysis Program-Transient (FRAP-T) computer code. These comparisons provide insight into both the experimental measurements methods and the validity of the gap conductance models

  11. Heavy-Section Steel Technology Program intermediate-scale pressure vessel tests

    International Nuclear Information System (INIS)

    Bryan, R.H.; Merkle, J.G.; Smith, G.C.; Whitman, G.D.

    1977-01-01

    The tests of intermediate-size vessels with sharp flaws permitted the comparison of experimentally observed behavior with analytical predictions of the behavior of flawed pressure vessels. Fracture strains estimated by linear elastic fracture mechanics (LEFM) were accurate in the cases in which the flaws resided in regions of high transverse restraint and the fracture toughness was sufficiently low for unstable fracture to occur prior to yielding through the vessel wall. When both of these conditions were not present, unstable fracture did occur, always preceded by stable crack growth; and the cylinders with flaws initially less than halfway through the wall attained gross yield prior to burst. Predictions of failure pressure of the vessels with flawed nozzles, based upon LEFM estimates of failure strain, were very conservative. LEFM calculations of critical load were based upon small-specimen fracture toughness test data. Whenever gross yielding preceded failure, the actual strains achieved were considerably greater than the estimated strains at failure based on LEFM. In such cases the strength of the vessel may be no longer dependent upon plane-strain fracture toughness but upon the capacity of the cracked section to carry the imposed load stably in the plastic range. Stable crack growth, which has not been predictable quantitatively, is an important factor in elastic-plastic analysis of strength. The ability of the flawed vessels to attain gross yield in unflawed sections has important qualitative implications on pressure vessel safety margins. The gross yield condition occurs in light-water-reactor pressure vessels at about 2 x design pressure. The intermediate vessel tests that demonstrated a capacity for exceeding this load confirm that the presumed margin of safety is not diminished by the presence of flaws of substantial size, provided that material properties are adequate

  12. Test of 6-inch-thick pressure vessels. Series 2. Intermediate test vessels V-3, V-4, and V-6

    International Nuclear Information System (INIS)

    Bryan, R.H.; Merkle, J.G.; Raftenberg, M.N.; Robinson, G.C.; Smith, J.E.

    1975-11-01

    The second series of intermediate vessel tests were crack initiation fracture tests of 6-in.-thick 39-in.-OD steel vessels with sharp surface flaws approximately 2 1 / 2 in. deep by 8 in. long in the longitudinal weld seams of the test cylinders. Fracture was initiated by means of hydraulic pressurization. One vessel was tested at each of three temperatures: 75, 130, and 190 0 F. Pretest analyses were made to predict the failure pressures and strains. Fracture toughness data obtained by equivalent-energy analysis of precracked Charpy-V tests and compact-tension specimen tests were used in the fracture analyses. The vessels behaved generally as had been expected. Posttest fracture analyses were also performed for each vessel. Detailed discussions of the fracture analysis methods developed in support of the vessel tests described are included. 34 references

  13. The need to pressure test prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Forgie, J.H.; Holland, J.A.

    1983-01-01

    In the period when PCRV were relatively unproven, proof pressure testing provided a useful demonstration of vessel integritiy and a confirmation of model testing and of analysis. No failures have occurred during concrete vessel tests in the UK or in the subsequent operational life of the vessels and much has been learned of their behaviour in service. The paper examines the advantages and disadvantages of proof testing PCRV in the light of the above increased knowledge of vessel performance. The paper draws attention to certain hypothetical loading cases that could be more onerous than the proof test and suggests that pressure testing could itself cause unnecessarily high loading to parts of the vessel. Always recognising the safety considerations and demonstrations of such are of prime importance, the authors suggest that a lower pressure level could be adopted without loss of original intent. In addition some ground rules are suggested as to cases where proof testing could be omitted. (orig./HP)

  14. Testing of Full Scale Flight Qualified Kevlar Composite Overwrapped Pressure Vessels

    Science.gov (United States)

    Greene, Nathanael; Saulsberry, Regor; Yoder, Tommy; Forsyth, Brad; Thesken, John; Phoenix, Leigh

    2007-01-01

    Many decades ago NASA identified a need for low-mass pressure vessels for carrying various fluids aboard rockets, spacecraft, and satellites. A pressure vessel design known as the composite overwrapped pressure vessel (COPV) was identified to provide a weight savings over traditional single-material pressure vessels typically made of metal and this technology has been in use for space flight applications since the 1970's. A typical vessel design consisted of a thin liner material, typically a metal, overwrapped with a continuous fiber yarn impregnated with epoxy. Most designs were such that the overwrapped fiber would carry a majority of load at normal operating pressures. The weight advantage for a COPV versus a traditional singlematerial pressure vessel contributed to widespread use of COPVs by NASA, the military, and industry. This technology is currently used for personal breathing supply storage, fuel storage for auto and mass transport vehicles and for various space flight and aircraft applications. The NASA Engineering and Safety Center (NESC) was recently asked to review the operation of Kevlar 2 and carbon COPVs to ensure they are safely operated on NASA space flight vehicles. A request was made to evaluate the life remaining on the Kevlar COPVs used on the Space Shuttle for helium and nitrogen storage. This paper provides a review of Kevlar COPV testing relevant to the NESC assessment. Also discussed are some key findings, observations, and recommendations that may be applicable to the COPV user community. Questions raised during the investigations have revealed the need for testing to better understand the stress rupture life and age life of COPVs. The focus of this paper is to describe burst testing of Kevlar COPVs that has been completed as a part of an the effort to evaluate the effects of ageing and shelf life on full scale COPVs. The test articles evaluated in this discussion had a diameter of 22 inches for S/N 014 and 40 inches for S/N 011. The

  15. V/V(max) test applied to SMM gamma-ray bursts

    Science.gov (United States)

    Matz, S. M.; Higdon, J. C.; Share, G. H.; Messina, D. C.; Iadicicco, A.

    1992-01-01

    We have applied the V/V(max) test to candidate gamma-ray bursts detected by the Gamma-Ray Spectrometer (GRS) aboard the SMM satellite to examine quantitatively the uniformity of the burst source population. For a sample of 132 candidate bursts identified in the GRS data by an automated search using a single uniform trigger criterion we find average V/V(max) = 0.40 +/- 0.025. This value is significantly different from 0.5, the average for a uniform distribution in space of the parent population of burst sources; however, the shape of the observed distribution of V/V(max) is unusual and our result conflicts with previous measurements. For these reasons we can currently draw no firm conclusion about the distribution of burst sources.

  16. R6 validation exercise: through thickness residual stress measurements on an experiment test vessel ring

    International Nuclear Information System (INIS)

    Mitchell, D.H.

    1988-06-01

    A series of bursting tests on thick-walled pressure vessels has been carried out as part of a validation exercise for the CEGB R6 failure assessment procedure. The objective of these tests was the examination of the behaviour of typical PWR primary vessel material subject to residual stresses in addition to primary loading with particular reference to the R6 assessment procedure. To this end, a semi-elliptic part-through defect was sited in the vessel longitudinal seam, which was a submerged arc weld in the non stress-relieved condition; it was then pressure tested to failure. Prior to the final assembly of this vessel, a ring of material was cut from it to act as a test-piece on which a residual stress survey could be made. Surface measurements using the centre-hole technique were made by CERL personnel, and this has been followed by two through- thickness measurements at BNL using the deep-hole technique. This paper describes these deep-hole measurements and presents the results from them. (author)

  17. Failure analysis of burst tested fuel tube samples

    International Nuclear Information System (INIS)

    Padmaprabu, C.; Ramana Rao, S.V.; Srivatsava, R.K.

    2005-01-01

    The Total Circumferential Elongation (TCE) is an important parameter for evaluation of ductility of the Zircaloy-4 fuel tubes for the PHWR reactors. The TCE values of the fuel tubes were obtained using the burst testing technique. In some lots there is a variation in the values of the TCE. To investigate the reasons for such a large variation in the TCE, samples were selected at appropriate intervals and sectioned at the fractured portion. The surface morphology of the fractured surfaces was examined under Scanning Electron Microscope (SEM) equipped with Energy Dispersive Spectrometer (EDS). The morphologies show segregation of elements at specific locations. Energy dispersive spectra was obtained from those segregated particles. According to the magnitude of TCE value the samples were classified into low, intermediate and high ductility. Low ductility samples were found to contain large amount of segregations along the thickness direction of the tube. This forms a brittle region and a path for the easy crack growth along thickness direction. In the case of intermediate samples the segregation occurred in fewer locations compared to low ductile samples and also confined to the circumferential direction of the outside surface of the tube. Due to this, probability of crack formation at the surface of the tube could be high. But crack growth would be slower in the ductile matrix along the thickness direction resulting in the enhancement of TCE value compared to the low ductile sample. In the high ductile samples, the segregations were very scarce and found to be isolated and embedded in the ductile matrix. The mode of failure in these types of samples was found to be purely ductile. Cracks were found to originate solely from the micro voids in the material. As the probability of crack formation and its propagation is low, very high TCE values were observed in these samples. Microstructural observations of fractured surfaces and EDAX analysis was able to identify the

  18. Heavy Section Steel Technology Program. Part II. Intermediate vessel testing

    International Nuclear Information System (INIS)

    Whitman, G.D.

    1975-01-01

    The testing of the intermediate pressure vessels is a major activity under the Heavy Section Steel Technology Program. A primary objective of these tests is to develop or verify methods of fracture prediction, through the testing of selected structures and materials, in order that a valid basis can be established for evaluating the serviceability and safety of light-water reactor pressure vessels. These vessel tests were planned with sufficiently specific objectives that substantial quantitative weight could be given to the results. Each set of testing conditions was chosen so as to provide specific data by which analytical methods of predicting flaw growth, and in some cases crack arrest, could be evaluated. Every practical effort was made to assure that results would be relevant to some aspect of real reactor pressure vessel performance through careful control of material properties, selection of test temperatures, and design of prepared flaws. 5 references

  19. Proving Test on the Reliability for Reactor Containment Vessel

    International Nuclear Information System (INIS)

    Takumi, K.; Nonaka, A.

    1988-01-01

    NUPEC (Nuclear Power Engineering Test Center) has started an eight-year project of Proving Test on the Reliability for Reactor Containment Vessel since June 1987. The objective of this project is to confirm the integrity of containment vessels under severe accident conditions. This paper shows the outline of this project. The test Items are (1) Hydrogen mixing and distribution test, (2) Hydrogen burning test, (3) Iodine trapping characteristics test, and (4) Structural behavior test. Based on the test results, computer codes are verified and as the results of analysis and evaluation by the computer codes, containment integrity is to be confirmed

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

    International Nuclear Information System (INIS)

    Ioan, M.; Hororoi, M.

    2013-01-01

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

  1. Instrumentation and testing of a prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Hessheimer, M.F.; Pace, D.W.; Klamerus, E.W.

    1997-01-01

    Static overpressurization tests of two scale models of nuclear containment structures - a steel containment vessel (SCV) representative of an improved, boiling water reactor (BWR) Mark II design and a prestressed concrete containment vessel (PCCV) for pressurized water reactors (PWR) - are being conducted by Sandia National Laboratories for the Nuclear Power Engineering Corporation of Japan and the U.S. Nuclear Regulatory Commission. This paper discusses plans for instrumentation and testing of the PCCV model. 6 refs., 2 figs., 2 tabs

  2. Burst Testing and Analysis of Superalloy Disks With a Dual Grain Microstructure

    Science.gov (United States)

    Gayda, John; Kantzos, Pete

    2006-01-01

    Elastic-plastic finite element analyses of room temperature burst tests on four superalloy disks were conducted and reported in this paper. Two alloys, Rene 104 (General Electric Aircraft Engines) and Alloy 10 (Honeywell Engines & Systems), were studied. For both alloys an advanced dual microstructure disk, fine grain bore and coarse grain rim, were analyzed and compared with conventional disks with uniform microstructures, coarse grain for Rene 104 and fine grain for Alloy 10. The analysis and experimental data were in good agreement up to burst. At burst, the analysis underestimated the speed and growth of the Rene 104 disks, but overestimated the speed and growth of the Alloy 10 disks. Fractography revealed that the Alloy 10 disks displayed significant surface microcracking and coalescence in comparison to Rene 104 disks. This phenomenon may help explain the differences between the Alloy 10 disks and the Rene 104 disks, as well as the observed deviations between analytical and experimental data at burst.

  3. Inductive testing of reactor pressure vessels

    International Nuclear Information System (INIS)

    Bergh, H.

    1987-01-01

    In Service Inspection of Reactor Pressure Vessels is mostly done with ultrasonics. Using special 2 crystal-probes good detectability is achieved for near surface defects. The problem is to detect closely spaced cracks, to decide if the defects are surface braking and, if not, to decide the remaining ligament. The purpose of this study is to investigate to what extent Eddy Current can solve these problems. Detecting surfacebreaking cracks and fields of cracks can be done using conventional Eddy Current techniques. Mapping of closely spaced cracks requires a small probe and a high frequency. Measurement of depths a larger probe, a lower frequency and knowledge of the crackfield since 2 closely spaced shallow cracks might be mistaken for one deep crack. Depths of singel cracks can be measured down to 7-8 mm. In closely spaced crackfields the depths can not be measured. The measurement is mostly based on amplitude. For not surface breaking defects the problem is to decide the ligament, i.e. the distance between surface and cracktip. To achieve good penetration a large probe, low frequency and high energy or pulsed energy is used. Ligament up to 4 mm can be measured with good accuracy. The measurements is mostly based on phase. Noise, which originates from rough surface, varied material structure and lift off, can be reduced using multi frequency mix, probe design and scanning pattern. (author)

  4. Seismic proving test of PWR reactor containment vessel

    International Nuclear Information System (INIS)

    Akiyama, H.; Yoshikawa, T.; Tokumaru, Y.

    1987-01-01

    The seismic reliability proving tests of nuclear power plant facilities are carried out by Nuclear Power Engineering Test Center (NUPEC), using the large-scale, high-performance vibration of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry (MITI). In 1982, the seismic reliability proving test of PWR containment vessel started using the test component of reduced scale 1/3.7 and the test component proved to have structural soundness against earthquakes. Subsequently, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. Whereupon, the seismic analysis and evaluation on the actual containment vessel were performed by these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed

  5. Acoustic emission monitoring of HFIR vessel during hydrostatic testing

    International Nuclear Information System (INIS)

    Friesel, M.A.; Dawson, J.F.

    1992-08-01

    This report discusses the results and conclusions reached from applying acoustic emission monitoring to surveillance of the High Flux Isotope Reactor vessel during pressure testing. The objective of the monitoring was to detect crack growth and/or fluid leakage should it occur during the pressure test. The report addresses the approach, acoustic emission instrumentation, installation, calibration, and test results

  6. Imperfection detection probability at ultrasonic testing of reactor vessels

    International Nuclear Information System (INIS)

    Kazinczy, F. de; Koernvik, L.Aa.

    1980-02-01

    The report is a lecture given at a symposium organized by the Swedish nuclear power inspectorate on February 1980. Equipments, calibration and testing procedures are reported. The estimation of defect detection probability for ultrasonic tests and the reliability of literature data are discussed. Practical testing of reactor vessels and welded joints are described. Swedish test procedures are compared with other countries. Series of test data for welded joints of the OKG-2 reactor are presented. Future recommendations for testing procedures are made. (GBn)

  7. Pre-service Acoustic Emission Testing for Metal Pressure Vessel

    International Nuclear Information System (INIS)

    Lee, Jong O; Yoon, Woon Ha; Lee, Tae Hee; Lee, Jong Kyu

    2003-01-01

    The field application of acoustic emission(AE) testing for brand-new metal pressure vessel were performed. We will introduce the test procedure for acoustic emission test such as instrument check distance between sensors, sensor location, whole system calibration, pressurization sequence, noise reduction and evaluation. The data of acoustic emission test contain many noise signal, these noise can be reduced by time filtering which based on the description of observation during AE test

  8. High temperature helium test rig with prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Schmidl, H.

    1975-10-01

    The report gives a short description of the joint project prestressed concrete vessel-helium test station as there is the building up of the concrete structure, the system of instrumentation, the data processing, the development of the helium components as well as the testing programs. (author)

  9. Pressure test method for reactor pressure vessel in construction field

    International Nuclear Information System (INIS)

    Takeda, Masakado; Ushiroda, Koichi; Miyahara, Ryohei; Takano, Hiroshi; Matsuura, Tadashi; Sato, Keiya.

    1998-01-01

    Plant constitutional parts as targets of both of a primary pressure test and a secondary pressure test are disposed in communication with a reactor pressure vessel, and a pressure of the primary pressure test is applied to the targets of both tests, so that the primary pressure test and the second pressure test are conducted together. Since the number of pressure tests can be reduced to promote construction, and the number of workers can also be reduced. A pressure exceeding the maximum pressure upon use is applied to the pressure vessel after disposing the incore structures, to continuously conduct the primary pressure test and the secondary pressure test joined together and an incore flowing test while closing the upper lid of the pressure vessel as it is in the construction field. The number of opening/closing of the upper lid upon conducting every test can be reduced, and since the pressure resistance test is conducted after arranging circumference conditions for the incore flowing test, the tests can be conducted collectively also in view of time. (N.H.)

  10. On the state of acoustic emission analysis in pressure vessel and model vessel testing

    International Nuclear Information System (INIS)

    Morgner, W.; Theis, K.; Henke, F.; Imhof, D.

    1985-01-01

    In the GDR acoustic emission analysis is being applied primarily in connection with hydraulic pressure testing of vessels in chemical industry. It is, however, also used for testing and monitoring of equipment and components in other branches of industry. The state-of-the-art is presented with regard to equipment needed, training of personnel, licensing of testing methods and appropriate testing procedures. In particular, the evaluation of the sum curves and amplitude distributions is explained, using rupture tests of two oxygen cylinders and a compressed-air bottle as examples. (author)

  11. Multilayer Pressure Vessel Materials Testing and Analysis. Phase 1

    Science.gov (United States)

    Cardinal, Joseph W.; Popelar, Carl F.; Page, Richard A.

    2014-01-01

    To provide NASA a comprehensive suite of materials strength, fracture toughness and crack growth rate test results for use in remaining life calculations for aging multilayer pressure vessels, Southwest Research Institute (R) (SwRI) was contracted in two phases to obtain relevant material property data from a representative vessel. This report describes Phase 1 of this effort which includes a preliminary material property assessment as well as a fractographic, fracture mechanics and fatigue crack growth analyses of an induced flaw in the outer shell of a representative multilayer vessel that was subjected to cyclic pressure test. SwRI performed this Phase 1 effort under contract to the Digital Wave Corporation in support of their contract to Jacobs ATOM for the NASA Ames Research Center.

  12. Test and evaluation of pressure vessel materials

    International Nuclear Information System (INIS)

    Choi, Sun Pil; Hong, Jun Hwa; Nho, Kye Hoe; Han, Dae June; Chi, Se Hwan

    1985-01-01

    We have prepared a method for analyzing the Charpy impact test data, which is deduced from ''the standard anelastic solid equation''. The theoretical expression for the absorbed energy is in a form of W=Wsub(U)+(Wsub(R)-Wsub(U))/ [1+(ωtau) 2 ] showing the Debye characteristics and where tau is given by the Arrhenius equation; tau=tau 0 exp(ΔH/ksub(B)T). Four measurable parameters, at the present stage, can characterize the dynamic hehavior of cracking (Charpy impact result). They are the upper shelf energy(Wsub(R), the lower shelf energy (Wsub(U)), the activation energy of crack (ΔH, and wtau(0) where w tau(0) are the resonance frequency of the specimen and the jumping pre-exponential factor of propagating crack respectively. However the states of R (relaxed) and U (un-relaxed) should be defined from reasonable physical conditions in the future and it is possible that Wsub(U) is small enough to be taken as zero. The effects of irradiation, alloying elements, and heat treatment on the impact results should be interpreted as changes in the above characteristic parameters. The present method has been applied for weld metal of SA 508-2 irradiated up to a fluence of 4x10 18 n/cm 2 , E>1.0Mev, resulting in about 29% decrease in Wsub(R), negligible change in Wsub(U), 5.6 times increase in ωtau 0 , and no change in ΔH. This seems to indicate that irradiation degrades an average value of YOUNG's modulus so that cracks propagate more easily and it does not effect on breaking the lattice bond. However much more systematic analyses should be necessary for correct judgment. It is concluded that the present method is quite adequate for analyzing the Charpy impact data even though plastic deformation in the specimen was not considered separately so that the method should be applied for various cases in order to evaluate the proper trend of effects of irradiation, alloying elements, and heat treatment on the Charpy impact results. (Author)

  13. Helium leak testing of large pressure vessels or subassemblies

    International Nuclear Information System (INIS)

    Hopkins, J.S.; Valania, J.J.

    1977-01-01

    Specifications for pressure-vessel components [such as the intermediate heat exchangers (IHX)] for service in the liquid metal fast breeder reactor facilities require helium leak testing of pressure boundaries to very exacting standards. The experience of Foster Wheeler Energy Corporation (FWEC) in successfully leak-testing the IHX shells and bundle assemblies now installed in the Fast Flux Test Facility at Richland, WA is described. Vessels of a somewhat smaller size for the closed loop heat exchanger system in the Fast Flux Test Facility have also been fabricated and helium leak tested for integrity of the pressure boundary by FWEC. Specifications on future components call for helium leak testing of the tube to tubesheet welds of the intermediate heat exchangers

  14. Dynamic testing of MFTF containment-vessel structural system

    International Nuclear Information System (INIS)

    Weaver, H.J.; McCallen, D.B.; Eli, M.W.

    1982-01-01

    Dynamic (modal) testing was performed on the Magnetic Fusion Test Facility (MFTF) containment vessel. The seismic design of this vessel was heavily dependent upon the value of structural damping used in the analysis. Typically for welded steel vessels, a value of 2 to 3% of critical is used. However, due to the large mass of the vessel and magnet supported inside, we felt that the interaction between the structure and its foundation would be enhanced. This would result in a larger value of damping because vibrational energy in the structure would be transferred through the foundation into the surrounding soil. The dynamic test performed on this structure (with the magnet in place) confirmed this later theory and resulted in damping values of approximately 4 to 5% for the whole body modes. This report presents a brief description of dynamic testing emphasizing the specific test procedure used on the MFTF-A system. It also presents an interpretation of the damping mechanisms observed (material and geometric) based upon the spatial characteristics of the modal parameters

  15. The emission of Gamma Ray Bursts as a test-bed for modified gravity

    Directory of Open Access Journals (Sweden)

    S. Capozziello

    2015-11-01

    Full Text Available The extreme physical conditions of Gamma Ray Bursts can constitute a useful observational laboratory to test theories of gravity where very high curvature regimes are involved. Here we propose a sort of curvature engine capable, in principle, of explaining the huge energy emission of Gamma Ray Bursts. Specifically, we investigate the emission of radiation by charged particles non-minimally coupled to the gravitational background where higher order curvature invariants are present. The coupling gives rise to an additional force inducing a non-geodesic motion of particles. This fact allows a strong emission of radiation by gravitationally accelerated particles. As we will show with some specific model, the energy emission is of the same order of magnitude of that characterizing the Gamma Ray Burst physics. Alternatively, strong curvature regimes can be considered as a natural mechanism for the generation of highly energetic astrophysical events. Possible applications to cosmology are discussed.

  16. Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

    International Nuclear Information System (INIS)

    Zado, V.

    1998-01-01

    The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

  17. Acoustic emission signal measurements in pressure vessel testing

    International Nuclear Information System (INIS)

    Peter, A.

    1984-01-01

    The number of acoustic emission events per plastically deformed unit of volume caused by artificial notches in real pressure vessels has been calculated taking into account reference voltage, distance between acoustic emission source and sensor as well as the effect of noise background. A test performed at a 100 m 3 gasholder verifies the theoretical considerations. (author)

  18. Manufacture, testing and assembly preparation of the JET vacuum vessel

    International Nuclear Information System (INIS)

    Arbez, J.; Baeumel, S.; Dean, J.R.; Duesling, G.; Froger, C.; Hemmerich, J.L.; Walravens, M.; Walter, K.; Winkel, T.

    1983-01-01

    To reach the target pressure of 10 -9 mbar, JET's double-walled Inconel vacuum vessel is being manufactured and assembled in clean conditions and with meticulous leak detection. Each octant (1/8 of the torus) is baked in an oven to 520 0 C and leak tested at 350 0 C to reveal leaks as small as 10 -9 mbar l/s, which are repaired. In service the vessel will be baked periodically to 500 0 C by CO 2 passing between its walls. The single-walled ports will be electrically heated. (author)

  19. Testing program for burning plasma experiment vacuum vessel bolted joint

    International Nuclear Information System (INIS)

    Hsueh, P.K.; Khan, M.Z.; Swanson, J.; Feng, T.; Dinkevich, S.; Warren, J.

    1992-01-01

    As presently designed, the Burning Plasma Experiment vacuum vessel will be segmentally fabricated and assembled by bolted joints in the field. Due to geometry constraints, most of the bolted joints have significant eccentricity which causes the joint behavior to be sensitive to joint clamping forces. Experience indicates that as a result of this eccentricity, the joint will tend to open at the side closest to the applied load with the extent of the opening being dependent on the initial preload. In this paper analytical models coupled with a confirmatory testing program are developed to investigate and predict the non-linear behavior of the vacuum vessel bolted joint

  20. The Assembly and Test of Pressure Vessel for Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kook Nam; Lee, Jong Min; Youn, Young Jung; June, Hyung Kil; Ahn, Sung Ho; Lee, Kee Hong; Kim, Young Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kennedy, Timothy C. [Oregon State University, Corvallis (United States)

    2009-02-15

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts: the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

  1. The Assembly and Test of Pressure Vessel for Irradiation

    International Nuclear Information System (INIS)

    Park, Kook Nam; Lee, Jong Min; Youn, Young Jung; June, Hyung Kil; Ahn, Sung Ho; Lee, Kee Hong; Kim, Young Ki; Kennedy, Timothy C.

    2009-01-01

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts: the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature

  2. Manipulator for testing a top-opened reactor pressure vessel

    International Nuclear Information System (INIS)

    Bauer, R.; Kastl, H.

    1991-01-01

    The design is described of a manipulator to be inserted into the inside of reactor pressure vessels opened at the top. The main components of the manipulator include a fixed column protruding into the pressure vessel and a support which is slidable on the column and carries the bearing component for the measuring, testing, inspection and repair instruments. The device includes a driving equipment for the support as well as the power supply for the sets accommodated on the support, with the aim to reduce the failure rate of the manipulator as a whole, shorten the time necessary for its assembling and thus the time of staying in the reactor pressure vessel and, at the same time, make its maintenance and operation easier. (Z.S.). 13 figs

  3. Multilayer Pressure Vessel Materials Testing and Analysis Phase 2

    Science.gov (United States)

    Popelar, Carl F.; Cardinal, Joseph W.

    2014-01-01

    To provide NASA with a suite of materials strength, fracture toughness and crack growth rate test results for use in remaining life calculations for the vessels described above, Southwest Research Institute® (SwRI®) was contracted in two phases to obtain relevant material property data from a representative vessel. An initial characterization of the strength, fracture and fatigue crack growth properties was performed in Phase 1. Based on the results and recommendations of Phase 1, a more extensive material property characterization effort was developed in this Phase 2 effort. This Phase 2 characterization included additional strength, fracture and fatigue crack growth of the multilayer vessel and head materials. In addition, some more limited characterization of the welds and heat affected zones (HAZs) were performed. This report

  4. Testing for lightning as a source of radio bursts observed on the nightside of Venus

    International Nuclear Information System (INIS)

    Sonwalkar, V.S.; Carpenter, D.L.; Strangeway, R.J.

    1990-11-01

    In certain previous studies of radio burst events recorded by the Pioneer Venus Orbiting Electric Field Detector (OEFD), data were sorted for statistical purposes according to occurrence at filter band frequencies smaller than or greater than typical values of the ambient electron gyrofrequency. The expectation in making this distinction was that the lowest frequency signals, at 100 Hz, were candidates for propagation through the ionosphere to the spacecraft in the whistler mode, and that the higher frequency signals, if of subionospheric origin, would require some different ionospheric penetration mechanism. On the basis of certain assumptions about the homogeneity and horizontal stratification of the Venusian nightside ionosphere, methods were developed for case-by-case testing of the hypothesis that any particular burst event originated in subionospheric lightning. The tests, which are capable of refinement, allow prediction of the resonance cone angle, refractive index, wave dispersion, and wave polarization. The tests have been applied to data from 11 periods along 7 orbits, and are believed to represent an improved way of categorizing OEFD burst data for purposes of investigating source/propagation mechanisms. Four of the five burst events that were not found consistent with the lightning hypothesis involved receptions at multiple OEFD filter band frequencies

  5. Vacuum vessel for the tandem Mirror Fusion Test Facility

    International Nuclear Information System (INIS)

    Gerich, J.W.

    1986-01-01

    In 1980, the US Department of Energy gave the Lawrence Livermore National Laboratory approval to design and build a tandem Mirror Fusion Test Facility (MFTF-B) to support the goals of the National Mirror Program. We designed the MFTF-B vacuum vessel both to maintain the required ultrahigh vacuum environment and to structurally support the 42 superconducting magnets plus auxiliary internal and external equipment. During our design work, we made extensive use of both simple and complex computer models to arrive at a cost-effective final configuration. As part of this work, we conducted a unique dynamic analysis to study the interaction of the 32,000-tonne concrete-shielding vault with the 2850-tonne vacuum vessel system. To maintain a vacuum of 2 x 10 -8 torr during the physics experiments inside the vessel, we designed a vacuum pumping system of enormous capacity. The vacuum vessel (4200-m 3 internal volume) has been fabricated and erected, and acceptance tests have been completed at the Livermore site. The rest of the machine has been assembled, and individual systems have been successfully checked. On October 1, 1985, we began a series of integrated engineering tests to verify the operation of all components as a complete system

  6. Rupture tests with reactor pressure vessel head models

    International Nuclear Information System (INIS)

    Talja, H.; Keinaenen, H.; Hosio, E.; Pankakoski, P.H.; Rahka, K.

    2003-01-01

    In the LISSAC project (LImit Strains in Severe ACcidents), partly funded by the EC Nuclear Fission and Safety Programme within the 5th Framework programme, an extensive experimental and computational research programme is conducted to study the stress state and size dependence of ultimate failure strains. The results are aimed especially to make the assessment of severe accident cases more realistic. For the experiments in the LISSAC project a block of material of the German Biblis C reactor pressure vessel was available. As part of the project, eight reactor pressure vessel head models from this material (22 NiMoCr 3 7) were tested up to rupture at VTT. The specimens were provided by Forschungszentrum Karlsruhe (FzK). These tests were performed under quasistatic pressure load at room temperature. Two specimens sizes were tested and in half of the tests the specimens contain holes describing the control rod penetrations of an actual reactor pressure vessel head. These specimens were equipped with an aluminium liner. All six tests with the smaller specimen size were conducted successfully. In the test with the large specimen with holes, the behaviour of the aluminium liner material proved to differ from those of the smaller ones. As a consequence the experiment ended at the failure of the liner. The specimen without holes yielded results that were in very good agreement with those from the small specimens. (author)

  7. Testing and Improving the Luminosity Relations for Gamma-Ray Bursts

    Science.gov (United States)

    Collazzi, Andrew C.

    2012-01-01

    Gamma Ray Bursts (GRBs) have several luminosity relations where a measurable property of a burst light curve or spectrum is correlated with the burst luminosity. These luminosity relations are calibrated for the fraction of bursts with spectroscopic redshifts and hence the known luminosities. GRBs have thus become known as a type of "standard candle” where standard candle is meant in the usual sense that luminosities can be derived from measurable properties of the bursts. GRBs can therefore be used for the same cosmology applications as Type Ia supernovae, including the construction of the Hubble Diagram and measuring massive star formation rate. The greatest disadvantage of using GRBs as standard candles is that their accuracy is lower than desired. With the recent advent of GRBs as a new standard candle, every effort must be made to test and improve the distance measures. Here, methods are employed to do just that. First, generalized forms of two tests are performed on the luminosity relations. All the luminosity relations pass one of these tests, and all but two pass the other. Even with this failure, redundancies in using multiple luminosity relations allows all the luminosity relations to retain value. Next, the "Firmani relation” is shown to have poorer accuracy than first advertised. It is also shown to be derivable from two other luminosity relations. For these reasons, the Firmani relation is useless for cosmology. The Amati relation is then revisited and shown to be an artifact of a combination of selection effects. Therefore, the Amati relation is also not good for cosmology. Fourthly, the systematic errors involved in measuring a luminosity indicator (Epeak) are measured. The result is an irreducible systematic error of 28%. Finally, the work concludes with a discussion about the impact of the work and the future of GRB luminosity relations.

  8. Preliminary results of steel containment vessel model test

    International Nuclear Information System (INIS)

    Matsumoto, T.; Komine, K.; Arai, S.

    1997-01-01

    A high pressure test of a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of a steel containment vessel (SCV), representing an improved boiling water reactor (BWR) Mark II containment, was conducted on December 11-12, 1996 at Sandia National Laboratories. This paper describes the preliminary results of the high pressure test. In addition, the preliminary post-test measurement data and the preliminary comparison of test data with pretest analysis predictions are also presented

  9. Power Burst Facility severe-fuel-damage test program

    International Nuclear Information System (INIS)

    McCardell, R.K.; MacDonald, P.E.

    1982-01-01

    As a result of the Three Mile Island Unit 2 (TMI-2) accident, the United States Nuclear Regulatory Commission (USNRC) has initiated a severe fuel damage research program to investigate fuel rod and core response, and fission product and hydrogen release and transport during degraded core cooling accidents. This paper presents a discussion of the expected benefits of the PBF severe fuel damage tests to the nuclear industry, a description of the first five planned experiments, the results of pretest analysis performed to predict the fuel bundle heatup for the first two experiments, and a discussion of Phase II severe fuel damage experiments. Modifications to the fission product detection system envisioned for the later experiments are also described

  10. Integrated leak rate test results of JOYO reactor containment vessel

    International Nuclear Information System (INIS)

    Tamura, M.; Endo, J.

    1982-02-01

    Integrated leak rate tests of JOYO after the reactor coolant system had been filled with sodium have been performed two times since 1978 (February 1978 and December 1979). The tests were conducted with the in-containment sodium systems, primary argon cover gas system and air conditioning systems operating. Both the absolute pressure method and the reference chamber method were employed during the test. The results of both tests confirmed the functioning of the containment vessel, and leak rate limits were satisfied. In Addition, the adequancy of the test instrumentation system and the test method was demonstrated. Finally the plant conditions required to maintain reasonable accuracy for the leak rate testing of LMFBR were established. In this paper, the test conditions and the test results are described. (author)

  11. 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)

  12. Testing of VVER reactor pressure vessels by TOFD method

    International Nuclear Information System (INIS)

    Skala, Z.; Vit, J.

    2002-01-01

    The Time of Flight Diffraction Method (TOFD) - one of the new testing methods capable to obtain the real dimensions of flaws - is presented in the paper.The laboratory experiments on samples with artificial flaws and samples with artificially prepared cracks confirmed the high accuracy of flaw through wall extent sizing by TOFD. This accuracy was confirmed by qualification of methods and systems used by Skoda JS for the in-service inspections of WWER 440 vessel circumferential weld. The qualification also confirmed the ability of TOFD to detect reliably flaws, which can are not reliably detected by standard pulse echo testing. Based on the result of experiments and qualification, the TOFD method shall be used routinely by Skoda JS for the inspection of vessel circumferential welds root area and for sizing of flaws exceeding the acceptance level

  13. Inspection device for external examination of pressure vessels, preferably for ultrasonic testing of reactor vessels

    International Nuclear Information System (INIS)

    Figlhuber, D.; Gallwas, J.; Weber, R.; Weber, J.

    1978-01-01

    The inspection device is placed in the annular gap between pressure vessel and biological shield of the BWR. In the annulus there is arranged at least one longitudinal rail which has got vertical guideways. Along it there can be moved on testing paths a manipulator with the ultrasonic search unit. The manipulator drive is outside of the inspection annulus. It is coupled to the manipulator by means of a tension member being guided over a reversing unit mounted at the upper end of the longitudinal rail. As a tension member there may be used a drag chain; the drive and the reversing unit are provided with corresponding chain wheels. (DG) [de

  14. Ultrasonic testing of electron beam closure weld on pressure vessel

    International Nuclear Information System (INIS)

    Andrews, R.W.

    1975-01-01

    One of the special products manufactured at the General Electric Neutron Devices Department (GEND) is a small stainless steel vessel designed to hold a component under high pressure for long periods. The vessel is a thick-walled cylinder with a threaded receptacle into which a plug is screwed and welded after receiving the unit to be tested. The test cavity is then pressurized through a small diameter opening in the bottom and that opening is welded closed. When x-ray inspection techniques did not reveal defective welds at the threaded plug in a pressured vessel, occasional ''leakers'' occurred. With normal equipment tolerances, the electron beam spike tends to wander from the desired path, particularly at the root of the weld. Ultrasonic techniques were used to successfully inspect the weld. The testing technique is based on the observation that ultrasonic energy is reflected from the unwelded screw threads and not from the regions where the threads are completely fused together by welding. Any gas pore or any threaded region outside the weld bead can produce an echo. The units are rotated while the ultrasonic transducer travels in a direction parallel to the axis of rotation and toward the welded end. This produces a helical scan which is converted to a two-dimensional presentation in which incomplete welds can be noted. (U.S.)

  15. Demonstration tests for manufacturing the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Shimizu, Katsusuke; Onozuka, Masanori; Usui, Yukinori; Urata, Kazuhiro; Tsujita, Yoshihiro; Nakahira, Masataka; Takeda, Nobukazu; Kakudate, Satoshi; Ohmori, Junji; Shibanuma, Kiyoshi

    2007-01-01

    Demonstration tests for manufacturing and assembly of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel have been conducted to confirm manufacturing and assembly process of the vacuum vessel (VV). The full-scale partial mock-up fabrication was planned and is in progress. The results will be available in the near future. Field-joint assembly procedure has been demonstrated using a test stand. Due to limited accessibility to the outer shell at the field joint, some operations, including alignment of the splice plates, field-joint welding, and examination, were found to be very difficult. In addition, a demonstration test on the selected back-seal structures was performed. It was found that the tested structures have insufficient sealing capabilities and need further improvement. The applicability of ultrasonic testing methods has been investigated. Although side drilled holes of 2.4 mm in diameter were detected, detection of the slit-type defects and defect characterization were found to be difficult. Feasibility test of liquid penetrant testing has revealed that the selected liquid penetrant testing (LPT) solutions have sufficient low outgas rates and are applicable to the VV

  16. Demonstration tests for manufacturing the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Katsusuke [Mitsubishi Heavy Industries, Ltd., Kobe Shipyard and Machinery Works, Wadasaki-cho 1-1-1, Hyogo-ku, Kobe 652-8585 (Japan)], E-mail: katsusuke_shimizu@mhi.co.jp; Onozuka, Masanori [Mitsubishi Heavy Industries, Ltd., Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan); Usui, Yukinori; Urata, Kazuhiro; Tsujita, Yoshihiro [Mitsubishi Heavy Industries, Ltd., Kobe Shipyard and Machinery Works, Wadasaki-cho 1-1-1, Hyogo-ku, Kobe 652-8585 (Japan); Nakahira, Masataka; Takeda, Nobukazu; Kakudate, Satoshi; Ohmori, Junji; Shibanuma, Kiyoshi [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan)

    2007-10-15

    Demonstration tests for manufacturing and assembly of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel have been conducted to confirm manufacturing and assembly process of the vacuum vessel (VV). The full-scale partial mock-up fabrication was planned and is in progress. The results will be available in the near future. Field-joint assembly procedure has been demonstrated using a test stand. Due to limited accessibility to the outer shell at the field joint, some operations, including alignment of the splice plates, field-joint welding, and examination, were found to be very difficult. In addition, a demonstration test on the selected back-seal structures was performed. It was found that the tested structures have insufficient sealing capabilities and need further improvement. The applicability of ultrasonic testing methods has been investigated. Although side drilled holes of 2.4 mm in diameter were detected, detection of the slit-type defects and defect characterization were found to be difficult. Feasibility test of liquid penetrant testing has revealed that the selected liquid penetrant testing (LPT) solutions have sufficient low outgas rates and are applicable to the VV.

  17. The technology development for surveillance test of reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Sun Phil; Park, Day Young; Choi, Kwen Jai

    1997-12-01

    Benchmark test was performed in accordance with the requirement of US NRC Reg. Guide DG-1053 for Kori unit-1 in order to determine best-estimated fast neutron fluence irradiated into reactor vessel. Since the uncertainty of radiation analysis comes from the calculation error due to neutron cross-section data, reactor core geometrical dimension, core source, mesh density, angular expansion and convergence criteria, evaluation of calculational uncertainty due to analytical method was performed in accordance with the regulatory guide and the proof was performed for entire analysis by comparing the measurement value obtained by neutron dosimetry located in surveillance capsule. Best-estimated neutron fluence in reactor vessel was calculated by bias factor, neutron flux measurement value/calculational value, from reanalysis result from previous 1st through 4th surveillance testing and finally fluence prediction was performed for the end of reactor life and the entire period of plant life extension. Pressurized thermal shock analysis was performed in accordance with 10 CFR 50.61 using the result of neutron fluence analysis in order to predict the life of reactor vessel material and the criteria of safe operation for Kori unit 1 was reestablished. (author). 55 refs., 55 figs.

  18. Hydride transport vessel vibration and shock test report

    Energy Technology Data Exchange (ETDEWEB)

    Tipton, D.G.

    1998-06-01

    Sandia National Laboratories performed vibration and shock testing on a Savannah River Hydride Transport Vessel (HTV) which is used for bulk shipments of tritium. This testing is required to qualify the HTV for transport in the H1616 shipping container. The main requirement for shipment in the H1616 is that the contents (in this case the HTV) have a tritium leak rate of less than 1x10{sup {minus}7} cc/sec after being subjected to shock and vibration normally incident to transport. Helium leak tests performed before and after the vibration and shock testing showed that the HTV remained leaktight under the specified conditions. This report documents the tests performed and the test results.

  19. Hydride transport vessel vibration and shock test report

    International Nuclear Information System (INIS)

    Tipton, D.G.

    1998-06-01

    Sandia National Laboratories performed vibration and shock testing on a Savannah River Hydride Transport Vessel (HTV) which is used for bulk shipments of tritium. This testing is required to qualify the HTV for transport in the H1616 shipping container. The main requirement for shipment in the H1616 is that the contents (in this case the HTV) have a tritium leak rate of less than 1x10 -7 cc/sec after being subjected to shock and vibration normally incident to transport. Helium leak tests performed before and after the vibration and shock testing showed that the HTV remained leaktight under the specified conditions. This report documents the tests performed and the test results

  20. Accelerated irradiation test of gundremmingen reactor vessel trepan material

    International Nuclear Information System (INIS)

    Hawthorne, J.R.

    1992-08-01

    Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279 degrees C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed

  1. Accelerated irradiation test of Gundremmingen reactor vessel trepan material

    Energy Technology Data Exchange (ETDEWEB)

    Hawthorne, J.R. [Materials Engineering Associates, Inc., Lanham, MD (United States)

    1992-08-01

    Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279{degrees}C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed.

  2. Burst and inter-burst duration statistics as empirical test of long-range memory in the financial markets

    Science.gov (United States)

    Gontis, V.; Kononovicius, A.

    2017-10-01

    We address the problem of long-range memory in the financial markets. There are two conceptually different ways to reproduce power-law decay of auto-correlation function: using fractional Brownian motion as well as non-linear stochastic differential equations. In this contribution we address this problem by analyzing empirical return and trading activity time series from the Forex. From the empirical time series we obtain probability density functions of burst and inter-burst duration. Our analysis reveals that the power-law exponents of the obtained probability density functions are close to 3 / 2, which is a characteristic feature of the one-dimensional stochastic processes. This is in a good agreement with earlier proposed model of absolute return based on the non-linear stochastic differential equations derived from the agent-based herding model.

  3. MELCOR modeling of the PBF [Power Burst Facility] Severe Fuel Damage Test 1-4

    International Nuclear Information System (INIS)

    Madni, I.K.

    1990-01-01

    This paper describes a MELCOR Version 1.8 simulation of the Power Burst Facility (PBF) Severe Fuel Damage (SFD) Test 1--4. The input data for the analysis were obtained from the Test Results Report and from SCDAP/RELAP5 input. Results are presented for the transient liquid level in the test bundle, clad temperatures, shroud temperatures, clad oxidation and hydrogen generation, bundle geometry changes, fission product release, and heat transfer to the bypass flow. Comparisons are made with experimental data and with SCDAP/RELAP5 calculations. 10 refs., 7 figs

  4. Multi-rod burst test under a loss-of coolant accident condition, (4)

    International Nuclear Information System (INIS)

    Otomo, Takashi; Hashimoto, Masao; Kawasaki, Satoru; Furuta, Teruo; Uetsuka, Hiroshi

    1983-06-01

    Multi-rod burst test of No.7808 bundle was performed in steam to estimate quantitative coolant flow channel restriction caused by the ballooning of zircaloy claddings in a fuel assembly during a LOCA transient in LWRs. The test was conducted under the condition that the initial internal pressure in each rod was 35kg/cm 2 (RT) and the heating rate was 9 0 C/s in steam with flow rate of 0.4g/cm 2 .min. The following results were obtained; (1) Maximum and burst pressures in rods were in the range 45 to 48kg/cm 2 and 41 to 45kg/cm 2 , respectively. The burst temperature of cladding were estimated to be 850 to 880 0 C. (2) Axial portions of tubes with greater than 34% strain were observed in the range 0 to 40mm in most rod. The mean length was 19mm in the bundle. (3) The degree of maximum increase in cross-sectional area is 54.2% in the bundle(7 x 7) and 66.9% in the internal rods(5 x 5). (4) Maximum channel area restriction was 40.5% in the bundle(7 x 7) and 51.4% in the internal rods(5 x 5). (author)

  5. Evaluation of effective energy deposition in test fuel during power burst experiment in NSRR

    International Nuclear Information System (INIS)

    Ohnishi, Nobuaki; Inabe, Teruo

    1982-01-01

    In an inpile experiment to study the fuel behavior under reactivity-initiated accident conditions, it is of great importance to understand the time-dependent characteristics of the energy deposited in the test fuel by burst power. The evaluation of the time-dependent energy deposition requires the knowledge of the fission rates and energy deposition per fission in the test fuel, both as a function of time. In the present work, the authors attempted to evaluate the relative fission rate change in the test fuel subjected to the power burst testing in the NSRR through the measurements and analyses of the fission power changes in the NSRR. Utilizing a micro fission chamber and a conventional larger fission chamber, they successfully measured the reactor fission power change ranging over a dozen of decades in magnitude and a thousand seconds in time. The measured power transient agreed quite well with calculated results. In addition, the time-dependent energy deposition per fission in the test fuel including the energy contribution from the driver core was analytically evaluated. The analyses indicate that the energy of about 175 MeV/fission is promptly deposited in the test fuel and that the additional energy of about 11 MeV is deposited afterwards. Finally the fractions of energy deposited in the test fuel until various times after power burst were determined by coupling the time-dependent relative fissions and energy deposition per fission in the test fuel. The prompt energy deposition ranges from about 50 to 80% of the total energy deposition for the reactivity insertion between 1.5 and 4.7 $, and the remaining is the delayed energy deposition. (author)

  6. Integral experiments on in-vessel coolability and vessel creep: results and analysis of the FOREVER-C1 test

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Nourgaliev, R.R.; Dinh, T.N.; Karbojian, A. [Division of Nuclear Power Safety, Royal Institute of Technology, Drottning Kristinas Vaeg., Stockholm (Sweden)

    1999-07-01

    This paper describes the FOREVER (Failure Of REactor VEssel Retention) experimental program, which is currently underway at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS). The objectives of the FOREVER experiments are to obtain data and develop validated models (i) on the melt coolability process inside the vessel, in the presence of water (in particular, on the efficacy of the postulated gap cooling to preclude vessel failure); and (ii) on the lower head failure due to the creep process in the absence of water inside and/or outside the lower head. The paper presents the experimental results and analysis of the first FOREVER-C1 test. During this experiment, the 1/10th scale pressure vessel, heated to about 900degC and pressurized to 26 bars, was subjected to creep deformation in a non-stop 24-hours test. The vessel wall displacement data clearly shows different stages of the vessel deformation due to thermal expansion, elastic, plastic and creep processes. The maximum displacement was observed at the lowermost region of the vessel lower plenum. Information on the FOREVER-C1 measured thermal characteristics and analysis of the observed thermal and structural behavior is presented. The coupled nature of thermal and mechanical processes, as well as the effect of other system conditions (such as depressurization) on the melt pool and vessel temperature responses are analyzed. (author)

  7. High speed cinematography of the initial break-point of latex condoms during the air burst test.

    Science.gov (United States)

    Stube, R; Voeller, B; Davidhazy, A

    1990-06-01

    High speed cinematography of latex condoms inflated to burst under standard (ISO) conditions reveals that rupture of the condom typically is initiated at a small focal point on the shank of the condom and then rapidly propagates throughout the condom's surface, often ending with partial or full severance of the condom at its point of attachment to the air burst instrument. This sequence of events is the reverse of that sometimes hypothesized to occur, where initiation of burst was considered to begin at the attachment point and to constitute a testing method artifact. This hypothesis of breakage at the attachment point, if true, would diminish the value of the air burst test as a standard for assessing manufacturing quality control as well as for condom strength measurements and comparisons.

  8. Integration test of ITER full-scale vacuum vessel sector

    International Nuclear Information System (INIS)

    Nakahira, M.; Koizumi, K.; Oka, K.

    2001-01-01

    The full-scale Sector Model Project, which was initiated in 1995 as one of the Large Seven R and D Projects, completed all R and D activities planned in the ITER-EDA period with the joint effort of the ITER Joint Central Team (JCT), the Japanese, the Russian Federation (RF) and the United States (US) Home Teams. The fabrication of a full-scale 18 toroidal sector, which is composed of two 9 sectors spliced at the port center, was successfully completed in September 1997 with the dimensional accuracy of ± 3 mm for the total height and total width. Both sectors were shipped to the test site in JAERI and the integration test was begun in October 1997. The integration test involves the adjustment of field joints, automatic Narrow Gap Tungsten Inert Gas (NG-TIG) welding of field joints with splice plates, and inspection of the joint by ultrasonic testing (UT), which are required for the initial assembly of ITER vacuum vessel. This first demonstration of field joint welding and performance test on the mechanical characteristics were completed in May 1998 and the all results obtained have satisfied the ITER design. In addition to these tests, the integration with the mid plane port extension fabricated by the Russian Home Team, and the cutting and re-welding test of field joints by using full-remotized welding and cutting system developed by the US Home Team, are planned as post EDA activities. (author)

  9. Integration test of ITER full-scale vacuum vessel sector

    International Nuclear Information System (INIS)

    Nakahira, M.; Koizumi, K.; Oka, K.

    1999-01-01

    The full-scale Sector Model Project, which was initiated in 1995 as one of the Large Seven ITER R and D Projects, completed all R and D activities planned in the ITER-EDA period with the joint effort of the ITER Joint Central Team (JCT), the Japanese, the Russian Federation (RF) and the United States (US) Home Teams. The fabrication of a full-scale 18 toroidal sector, which is composed of two 9 sectors spliced at the port center, was successfully completed in September 1997 with the dimensional accuracy of - 3 mm for the total height and total width. Both sectors were shipped to the test site in JAERI and the integration test was begun in October 1997. The integration test involves the adjustment of field joints, automatic Narrow Gap Tungsten Inert Gas (NG-TIG) welding of field joints with splice plates, and inspection of the joint by ultrasonic testing (UT), which are required for the initial assembly of ITER vacuum vessel. This first demonstration of field joint welding and performance test on the mechanical characteristics were completed in May 1998 and the all results obtained have satisfied the ITER design. In addition to these tests, the integration with the mid plane port extension fabricated by the Russian Home Team, and the cutting and re-welding test of field joints by using full-remotized welding and cutting system developed by the US Home Team, are planned as post EDA activities. (author)

  10. Tone burst generator for a Non-Destructive Testing system based on ultrasonic guided waves

    OpenAIRE

    Jiménez Sánchez, Daniel

    2011-01-01

    English: This PFC provides a design of a tested and specific tone-burst generator circuit for a Non-Destructive System based on ultrasonid guided waves. This circuit includes a complementary protection circuit for the NDT system working in a pulse-echo configuration. In this paper, a brief state f art about different driving circuits employed in distinct NDE systems is presented. Castellano: El PFC proporciona un diseño electrónico específico y probado de un circuito excitador de salvas (C...

  11. A Theoretical Investigation of Composite Overwrapped Pressure Vessel (COPV) Mechanics Applied to NASA Full Scale Tests

    Science.gov (United States)

    Thesken, John C.; Murthy, Pappu L. N.; Phoenix, S. L.; Greene, N.; Palko, Joseph L.; Eldridge, Jeffrey; Sutter, James; Saulsberry, R.; Beeson, H.

    2009-01-01

    A theoretical investigation of the factors controlling the stress rupture life of the National Aeronautics and Space Administration's (NASA) composite overwrapped pressure vessels (COPVs) continues. Kevlar (DuPont) fiber overwrapped tanks are of particular concern due to their long usage and the poorly understood stress rupture process in Kevlar filaments. Existing long term data show that the rupture process is a function of stress, temperature and time. However due to the presence of a load sharing liner, the manufacturing induced residual stresses and the complex mechanical response, the state of actual fiber stress in flight hardware and test articles is not clearly known. This paper is a companion to a previously reported experimental investigation and develops a theoretical framework necessary to design full-scale pathfinder experiments and accurately interpret the experimentally observed deformation and failure mechanisms leading up to static burst in COPVs. The fundamental mechanical response of COPVs is described using linear elasticity and thin shell theory and discussed in comparison to existing experimental observations. These comparisons reveal discrepancies between physical data and the current analytical results and suggest that the vessel s residual stress state and the spatial stress distribution as a function of pressure may be completely different from predictions based upon existing linear elastic analyses. The 3D elasticity of transversely isotropic spherical shells demonstrates that an overly compliant transverse stiffness relative to membrane stiffness can account for some of this by shifting a thin shell problem well into the realm of thick shell response. The use of calibration procedures are demonstrated as calibrated thin shell model results and finite element results are shown to be in good agreement with the experimental results. The successes reported here have lead to continuing work with full scale testing of larger NASA COPV

  12. Engineering Evaluation/Cost Analysis for Power Burst Facility (PER-620) Final End State and PBF Vessel Disposal

    Energy Technology Data Exchange (ETDEWEB)

    B. C. Culp

    2007-05-01

    Preparation of this engineering evaluation/cost analysis is consistent with the joint U.S. Department of Energy and U.S. Environmental Protection Agency Policy on Decommissioning of Department of Energy Facilities Under the Comprehensive Environmental Response, Compensation, and Liability Act, (DOE and EPA 1995) which establishes the Comprehensive Environmental, Response, Compensation, and Liability Act non-time critical removal action process as an approach for decommissioning. The scope of this engineering evaluation/cost analysis is to evaluate alternatives and recommend a preferred alternative for the final end state of the PBF and the final disposal location for the PBF vessel.

  13. DELPHIN - a new system for testing reactor pressure vessels

    International Nuclear Information System (INIS)

    Dressler, K.

    1999-01-01

    The author discusses the question of whether a new concept for testing RPVs is necessary. He concentrates his exposition upon the RPV testing system DELPHIN recently developed by ABB, which has been successfully employed in vessel-interior tests since 1998. The new system reduces both the time required and the financial costs for RPV tests. The tests have become more efficient particularly as a result of new developments in the fields of handling machinery and microelectronics. As an example of the improved quality, the author quotes the ultrasonic system ZAQUS: thanks to the high quality of the ultrasonic data, rapid comparison with the results of earlier repreated tests on the RPV is now possible. Since problems of interpretation did not arise, the overall results in an initial application were available only two hours after the last data recording. The author's verdict: DELPHIN has successfully undergone its 'baptism by fire'; still in need of improvement, he states, is the occupany time of the pond, which is not yet as short as targeted. (orig.) [de

  14. Testing gravitational parity violation with coincident gravitational waves and short gamma-ray bursts

    International Nuclear Information System (INIS)

    Yunes, Nicolas; O'Shaughnessy, Richard; Owen, Benjamin J.; Alexander, Stephon

    2010-01-01

    Gravitational parity violation is a possibility motivated by particle physics, string theory, and loop quantum gravity. One effect of it is amplitude birefringence of gravitational waves, whereby left and right circularly polarized waves propagate at the same speed but with different amplitude evolution. Here we propose a test of this effect through coincident observations of gravitational waves and short gamma-ray bursts from binary mergers involving neutron stars. Such gravitational waves are highly left or right circularly polarized due to the geometry of the merger. Using localization information from the gamma-ray burst, ground-based gravitational wave detectors can measure the distance to the source with reasonable accuracy. An electromagnetic determination of the redshift from an afterglow or host galaxy yields an independent measure of this distance. Gravitational parity violation would manifest itself as a discrepancy between these two distance measurements. We exemplify such a test by considering one specific effective theory that leads to such gravitational parity violation, Chern-Simons gravity. We show that the advanced LIGO-Virgo network and all-sky gamma-ray telescopes can be sensitive to the propagating sector of Chern-Simons gravitational parity violation to a level roughly 2 orders of magnitude better than current stationary constraints from the LAGEOS satellites.

  15. Miniaturized Charpy test for reactor pressure vessel embrittlement characterization

    Energy Technology Data Exchange (ETDEWEB)

    Manahan, M.P. Sr. [MPM Research and Consulting, Lemont, PA (United States)

    1999-10-01

    Modifications were made to a conventional Charpy machine to accommodate the miniaturized Charpy V-Notch (MCVN) specimens which were fabricated from an archived reactor pressure vessel (RPV) steel. Over 100 dynamic MCVN tests were performed and compared to the results from conventional Charpy V-Notch (CVN) tests to demonstrate the efficacy of the miniature specimen test. The optimized sidegrooved MCVN specimens exhibit transitional fracture behavior over essentially the same temperature range as the CVN specimens which indicates that the stress fields in the MCVN specimens reasonably simulate those of the CVN specimens and this fact has been observed in finite element calculations. This result demonstrates a significant breakthrough since it is now possible to measure the ductile-brittle transition temperature (DBTT) using miniature specimens with only small correction factors, and for some materials as in the present study, without the need for any correction factor at all. This development simplifies data interpretation and will facilitate future regulatory acceptance. The non-sidegrooved specimens yield energy-temperature data which is significantly shifted downward in temperature (non-conservative) as a result of the loss of constraint which accompanies size reduction.

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

    International Nuclear Information System (INIS)

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

    1990-12-01

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

  17. Testing the Isotropic Universe Using the Gamma-Ray Burst Data of Fermi/GBM

    Science.gov (United States)

    Řípa, Jakub; Shafieloo, Arman

    2017-12-01

    The sky distribution of gamma-ray bursts (GRBs) has been intensively studied by various groups for more than two decades. Most of these studies test the isotropy of GRBs based on their sky number density distribution. In this work, we propose an approach to test the isotropy of the universe through inspecting the isotropy of the properties of GRBs such as their duration, fluences, and peak fluxes at various energy bands and different timescales. We apply this method on the Fermi/Gamma-ray Burst Monitor (GBM) data sample containing 1591 GRBs. The most noticeable feature we found is near the Galactic coordinates l≈ 30^\\circ , b≈ 15^\\circ , and radius r≈ 20^\\circ {--}40^\\circ . The inferred probability for the occurrence of such an anisotropic signal (in a random isotropic sample) is derived to be less than a percent in some of the tests while the other tests give results consistent with isotropy. These are based on the comparison of the results from the real data with the randomly shuffled data samples. Considering the large number of statistics we used in this work (some of which are correlated with each other), we can anticipate that the detected feature could be a result of statistical fluctuations. Moreover, we noticed a considerably low number of GRBs in this particular patch, which might be due to some instrumentation or observational effects that can consequently affect our statistics through some systematics. Further investigation is highly desirable in order to clarify this result, e.g., utilizing a larger future Fermi/GBM data sample as well as data samples of other GRB missions and also looking for possible systematics.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  19. Development of automatic ultrasonic testing equipment for reactor pressure vessel

    International Nuclear Information System (INIS)

    Jang, Kee Ok; Park, Dae Yung; Park, Moon Hoh; Koo, Kil Mo; Park, Kwang Heui; Kang, Sang Sin; Bang, Heui Song; Noh, Heui Choong; Kong, Woon Sik

    1994-08-01

    The selected weld areas of reactor pressure vessel and adjacent piping are examined by remote mechanized ultrasonic testing(MUT) equipment. Since the MUT equipment was purchased from Southwest Research Institute (SwRI) in April 1985, we have performed 15 inservice inspections and 5 preservice inspections. However, the reliability of examination was recently decreased rapidly as the problems which results from the old age of equipment and the frequent movement to plant site to site have occurred frequently. Therefore, the 3-axis control system hardware in occurring many problems among the equipments of mechanized ultrasonic testing (MUT) was designed and developed to cover the examination areas of nozzle-shell weld as specified in ASME Code Section XI and to improve the examination reliability. The new 3-axis control system hardware with the performance of this project was developed to be compatible with the old one and it was used as dual system or spare parts of the old system. Furthermore, the established technologies are expected to be applied to the similar control systems in nuclear power plant. 17 figs, 2 pix, 2 tabs, 10 refs. (Author)

  20. Development of automatic ultrasonic testing equipment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Kee Ok; Park, Dae Yung; Park, Moon Hoh; Koo, Kil Mo; Park, Kwang Heui; Kang, Sang Sin; Bang, Heui Song; Noh, Heui Choong; Kong, Woon Sik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-08-01

    The selected weld areas of reactor pressure vessel and adjacent piping are examined by remote mechanized ultrasonic testing(MUT) equipment. Since the MUT equipment was purchased from Southwest Research Institute (SwRI) in April 1985, we have performed 15 inservice inspections and 5 preservice inspections. However, the reliability of examination was recently decreased rapidly as the problems which results from the old age of equipment and the frequent movement to plant site to site have occurred frequently. Therefore, the 3-axis control system hardware in occurring many problems among the equipments of mechanized ultrasonic testing (MUT) was designed and developed to cover the examination areas of nozzle-shell weld as specified in ASME Code Section XI and to improve the examination reliability. The new 3-axis control system hardware with the performance of this project was developed to be compatible with the old one and it was used as dual system or spare parts of the old system. Furthermore, the established technologies are expected to be applied to the similar control systems in nuclear power plant. 17 figs, 2 pix, 2 tabs, 10 refs. (Author).

  1. Development of automatic Ultrasonic testing equipment for reactor pressure vessel

    International Nuclear Information System (INIS)

    Kim, Kor R.; Kim, Jae H.; Lee, Jae C.

    1996-06-01

    The selected weld areas of a reactor pressure vessel and adjacent piping are examined by the remote mechanized ultrasonic testing (MUT) equipment. Since the MUT equipment was purchased from southwest Research Institute (SwRI) in April 1985, 15 inservice inspections and 5 preservice inspections are performed with this MUT equipment. However due to the old age of the equipment and frequent movements to plant sites, the reliability of examination was recently decreased rapidly and it is very difficult to keep spare parts. In order to resolve these problems and to meet the strong request from plant sites, we intend to develop a new 3-axis control system including hardware and software. With this control system, we expect more efficient and reliable examination of the nozzle to shell weld areas, which is specified in ASME Code Section XI. The new 3-axis control system hardware and software were designed and development of our own control system, the advanced technologies of computer control mechanism were established and examination reliability of the nozzle to shell weld area was improved. With the development of our 3-axis control system for PaR ISI-2 computer control system, the reliability of nozzle to shell weld area examination has been improved. The established technologies from the development and detailed analysis of existing control system, are expected to be applied to the similar control systems in nuclear power plants. (author). 12 refs., 4 tabs., 33 figs

  2. Development of automatic Ultrasonic testing equipment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kor R.; Kim, Jae H.; Lee, Jae C.

    1996-06-01

    The selected weld areas of a reactor pressure vessel and adjacent piping are examined by the remote mechanized ultrasonic testing (MUT) equipment. Since the MUT equipment was purchased from southwest Research Institute (SwRI) in April 1985, 15 inservice inspections and 5 preservice inspections are performed with this MUT equipment. However due to the old age of the equipment and frequent movements to plant sites, the reliability of examination was recently decreased rapidly and it is very difficult to keep spare parts. In order to resolve these problems and to meet the strong request from plant sites, we intend to develop a new 3-axis control system including hardware and software. With this control system, we expect more efficient and reliable examination of the nozzle to shell weld areas, which is specified in ASME Code Section XI. The new 3-axis control system hardware and software were designed and development of our own control system, the advanced technologies of computer control mechanism were established and examination reliability of the nozzle to shell weld area was improved. With the development of our 3-axis control system for PaR ISI-2 computer control system, the reliability of nozzle to shell weld area examination has been improved. The established technologies from the development and detailed analysis of existing control system, are expected to be applied to the similar control systems in nuclear power plants. (author). 12 refs., 4 tabs., 33 figs.

  3. Adjustable guide for a testing system for reactor pressure vessels

    International Nuclear Information System (INIS)

    Seifert, W.

    1980-01-01

    The device consisting of a guide rail and a manipulator is introduced into the gap between pressure vessel wall and biological shield by means of suspending wire drums and manipulator drums. For adjustment of the device an elbow telescope is used. The guide rail is fixed to the pressure vessel wall by means of electromagnets. The movements of the manipulator with respect to the guide rail are performed with the aid of a motor. (DG) [de

  4. Strain measurement in and analysis for hydraulic test of CPR1000 reactor pressure vessel

    International Nuclear Information System (INIS)

    Zhou Dan; Zhuang Dongzhen

    2013-01-01

    The strain measurement in hydraulic test of CPR1000 reactor pressure vessel performed in Dongfang Heavy Machinery Co., Ltd. is introduced. The detail test scheme and method was introduced and the measurement results of strain and stress was given. Meanwhile the finite element analysis was performed for the pressure vessel, which was generally matched with the measurement results. The reliability of strain measurement was verified and the high strength margin of vessel was shown, which would give a good reference value for the follow-up hydraulic tests and strength analysis of reactor pressure vessel. (authors)

  5. Test facility for fast gas injections into a vessel filled with water

    International Nuclear Information System (INIS)

    Wilhelm, D.; Kirstahler, M.

    1987-11-01

    The Fast Gas Injection Facility (SGI) was set up to study the hydrodynamics during the expansion of a gas bubble into a vessel filled with water. The gas stored in a pressure vessel expands against gravity through a circular duct into a large cylindrical vessel partly with water. This report covers the description of the test facility and the data acquisition. Results of the first test series are added. (orig.) [de

  6. Testing the anisotropy in the angular distribution of Fermi/GBM gamma-ray bursts

    Science.gov (United States)

    Tarnopolski, M.

    2017-12-01

    Gamma-ray bursts (GRBs) were confirmed to be of extragalactic origin due to their isotropic angular distribution, combined with the fact that they exhibited an intensity distribution that deviated strongly from the -3/2 power law. This finding was later confirmed with the first redshift, equal to at least z = 0.835, measured for GRB970508. Despite this result, the data from CGRO/BATSE and Swift/BAT indicate that long GRBs are indeed distributed isotropically, but the distribution of short GRBs is anisotropic. Fermi/GBM has detected 1669 GRBs up to date, and their sky distribution is examined in this paper. A number of statistical tests are applied: nearest neighbour analysis, fractal dimension, dipole and quadrupole moments of the distribution function decomposed into spherical harmonics, binomial test and the two-point angular correlation function. Monte Carlo benchmark testing of each test is performed in order to evaluate its reliability. It is found that short GRBs are distributed anisotropically in the sky, and long ones have an isotropic distribution. The probability that these results are not a chance occurrence is equal to at least 99.98 per cent and 30.68 per cent for short and long GRBs, respectively. The cosmological context of this finding and its relation to large-scale structures is discussed.

  7. Heissdampfreaktor (HDR) steel-containment-vessel and floodwater-storage-tank structural-dynamics tests

    International Nuclear Information System (INIS)

    Arendts, J.G.

    1982-01-01

    Inertance (vibration) testing of two significant vessels at the Heissdampfreaktor (HDR) facility, located near Kahl, West Germany, was recently completed. Transfer functions were obtained for determination of the modal properties (frequencies, mode shapes and damping) of the vessels using two different test methods for comparative purposes. One of the vessels tested was the steel containment vessel (SCV). The SCV is approximately 180 feet high and 65 feet in diameter with a 1.2-inch wall thickness. The other vessel, called the floodwater storage tank (FWST), is a vertically standing vessel approximately 40 feet high and 10 feet in diameter with a 1/2-inch wall thickness. The FWST support skirt is square (in plan views) with its corners intersecting the ellipsoidal bottom head near the knuckle region

  8. Acoustic emission for interlaminar toughness testing of CFRP: Evaluation of the crack growth due to burst analysis

    Czech Academy of Sciences Publication Activity Database

    Lissek, F.; Haegerb, A.; Knoblauch, V.; Hloch, Sergej; Pude, F.; Kaufeld, M.

    2018-01-01

    Roč. 136, č. 1 (2018), s. 55-62 ISSN 1359-8368 Institutional support: RVO:68145535 Keywords : DCB * interlaminar toughness testing * acoustic emission * CFRP * burst analysis Subject RIV: JQ - Machines ; Tools Impact factor: 4.727, year: 2016 http://www.sciencedirect.com/science/article/pii/S1359836817313720

  9. Acoustic emission for interlaminar toughness testing of CFRP: Evaluation of the crack growth due to burst analysis

    Czech Academy of Sciences Publication Activity Database

    Lissek, F.; Haegerb, A.; Knoblauch, V.; Hloch, Sergej; Pude, F.; Kaufeld, M.

    2018-01-01

    Roč. 136, č. 1 (2018), s. 55-62 ISSN 1359-8368 Institutional support: RVO:68145535 Keywords : DCB * interlaminar toughness testing * acoustic emission * CFRP * burst analysis Subject RIV: JQ - Machines ; Tools Impact factor: 4.727, year: 2016 http://www. science direct.com/ science /article/pii/S1359836817313720

  10. Test of 6-in.-thick pressure vessels. Series 4: intermediate test vessels V-5 and V-9 with inside nozzle corner cracks

    International Nuclear Information System (INIS)

    Merkle, J.G.; Robinson, G.C.; Holz, P.P.; Smith, J.E.

    1977-01-01

    Failure testing is described for two 99-cm-diam (39-in.), 15.2-cm-thick (6-in.) steel pressure vessels, each containing one flawed nozzle. Vessel V-5 was tested at 88 0 C (190 0 F) and failed by leaking without fracturing after extensive stable crack growth. Vessel V-9 was tested at 25 0 C (75 0 F) and failed by fracturing. Material properties measured before the tests were used for pretest and posttest fracture analyses. Test results supported by analysis indicate that inside nozzle corner cracks are not subject to plane strain under pressure loading. The preparation of inside nozzle corner cracks is described in detail. Extensive experimental data are tabulated and plotted

  11. Observational tests of the Electro-Magnetic Black Hole Theory in Gamma-Ray Bursts

    OpenAIRE

    Ruffini, Remo

    2002-01-01

    The Relative Space-Time Transformation (RSTT) Paradigm and the Interpretation of the Burst Structure (IBS) Paradigm are applied to the analysis of the structure of the burst and afterglow of Gamma-Ray Bursts within the theory based on the vacuum polarization process occurring in an Electro-Magnetic Black Hole, the EMBH theory. This framework is applied to the study of the GRB991216 which is used as a prototype. The GRB-Supernova Time Sequence (GSTS) Paradigm, which introduces the concept of i...

  12. Fuel-coolant interaction visualization test for in-vessel corium retention external reactor vessel cooling (IVR-ERVC) condition

    Energy Technology Data Exchange (ETDEWEB)

    Na, Young Su; Hong, Seong Ho; Song, Jin Ho; Hong, Seong Wan [Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.

  13. Blowdown mass flow measurements during the Power Burst Facility LOC-11C test

    International Nuclear Information System (INIS)

    Broughton, J.M.; MacDonald, P.E.

    1979-01-01

    An interpretation and evaluation of the two-phase coolant mass flow measurements obtained during Test LOC-11C performed in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL) are presented. Although a density gradient existed within the pipe between 1 and 6 s, the homogeneous flow model used to calculate the coolant mass flow from the measured mixture density, momentum flux, and volumetric flow was found to be generally satisfactory. A cross-sectional average density was determined by fitting a linear density gradient through the upper and lower chordal densities obtained from a three-beam gamma densitometer and then combining the result with the middle beam density. The integrated measured coolant mass flow was subsequently found to be within 5% if the initial mass inventory of the PBF loss-of-coolant accident (LOCA) system. The posttest calculations using the RELAP4/MOD6 computer code to determine coolant mass flow for Test LOC-11C also agreed well with the measured data

  14. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    International Nuclear Information System (INIS)

    Biemiller, E.C.; Fyfitch, Stephen; Campbell, C.A.

    1994-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 982 o C (1800 o F)) and to simulate Regulatory Guide 1.99 database materials (austenitized at 871 o C (1600 o F)). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (University of Michigan Test Reactor) which had never been used before for this type of irradiation program. Materials taken from plate surface locations (versus 1/4 T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, are maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (260 o C and 288 o C) to determine the effect of irradiation temperature on embrittlement. (Author)

  15. Using An Adapter To Perform The Chalfant-Style Containment Vessel Periodic Maintenance Leak Rate Test

    International Nuclear Information System (INIS)

    Loftin, B.; Abramczyk, G.; Trapp, D.

    2011-01-01

    Recently the Packaging Technology and Pressurized Systems (PT and PS) organization at the Savannah River National Laboratory was asked to develop an adapter for performing the leak-rate test of a Chalfant-style containment vessel. The PT and PS organization collaborated with designers at the Department of Energy's Pantex Plant to develop the adapter currently in use for performing the leak-rate testing on the containment vessels. This paper will give the history of leak-rate testing of the Chalfant-style containment vessels, discuss the design concept for the adapter, give an overview of the design, and will present results of the testing done using the adapter.

  16. Benchmark of the SixTrack-Fluka Active Coupling Against the SPS Scrapers Burst Test

    CERN Multimedia

    Mereghetti, A; Cerutti, F

    2014-01-01

    The SPS scrapers are a key ingredient for the clean injection into the LHC: they cut off halo particles quite close to the beam core (e.g.~3.5 sigma) just before extraction, to minimise the risk for quenches. The improved beam parameters as envisaged by the LHC Injectors Upgrade (LIU) Project required a revision of the present system, to assess its suitability and robustness. In particular, a burst (i.e. endurance) test of the scraper blades has been carried out, with the whole bunch train being scraped at the centre (worst working conditions). In order to take into account the effect of betatron and longitudinal beam dynamics on energy deposition patterns, and nuclear and Coulomb scattering in the absorbing medium onto loss patterns, the SixTrack and Fluka codes have been coupled, profiting from the best of the refined physical models they respectively embed. The coupling envisages an active exchange of tracked particles between the two codes at each turn, and an on-line aperture check in SixTrack, in order ...

  17. Data on test results of vessel cooling system of high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

    2003-02-01

    High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

  18. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    International Nuclear Information System (INIS)

    Biemiller, E.C.; Fyfitch, S.; Campbell, C.A.

    1993-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 1800 deg F) and to simulate Regulatory Guide 1.99 database materials (austenitized at 1600 deg. F). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (U. of Michigan Test Reactor) which had never been used for this type of irradiation program. Materials taken from plate surface locations (vs. 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, is maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (500 deg. F and 550 deg. F) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an increase in T 30 shift of 69 deg. F for a decrease in irradiation temperature of 50 deg. F. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel had no apparent effect on irradiation response. No apparent microstructure

  19. Product consistency testing of three reference glasses in stainless steel and perfluoroalkoxy resin vessels

    International Nuclear Information System (INIS)

    Olson, K.M.; Smith, G.L.; Marschman, S.C.

    1995-03-01

    Because of their chemical durability, silicate glasses have been proposed and researched since the mid-1950s as a medium for incorporating high-level radioactive waste (HLW) generated from processing of nuclear materials. A number of different waste forms were evaluated and ranked in the early 1980s; durability (leach resistance) was the highest weighted factor. Borosilicate glass was rated the best waste form available for incorporation of HLW. Four different types of vessels and three different glasses were used to study the possible effect of vessel composition on durability test results from the Production Consistency Test (PCT). The vessels were 45-m 304 stainless steel vessels, 150-m 304 L stainless steel vessels, and 60-m perfluoroalkoxy (PFA) fluoropolymer resin vessels. The three glasses were the Environmental Assessment glass manufactured by Corning Incorporated and supplied by Westinghouse Savannah River company, and West Valley Nuclear Services reference glasses 5 and 6, manufactured and supplied by Catholic University of America. Within experimental error, no differences were found in durability test results using the 3 different glasses in the 304L stainless steel or PFA fluoropolymer resin vessels over the seven-day test period

  20. More recent developments for the ultrasonic testing of light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Seiger, H.; Engl, G.

    1976-01-01

    The development of an ultrasonic testing method for the inspection from the outside of the areas close to the cladding of the spherical fields of holes of light water reactor pressure vessels is described

  1. Acoustic emission test on a 25mm thick mild steel pressure vessel with inserted defects

    International Nuclear Information System (INIS)

    Bentley, P.G.; Dawson, D.G.; Hanley, D.J.; Kirby, N.

    1976-12-01

    Acoustic emission measurements have been taken on an experimental mild steel vessel with 4 inserted defects ranging in severity up to 90% of through thickness. The vessel was subjected to a series of pressure excursions of increasing magnitude until failure occurred by extension of the largest inserted defect through the vessel wall. No acoustic emission was detected throughout any part of the tests which would indicate the presence of such serious defects or of impending failure. Measurements of acoustic emission from metallurgical specimens are included and the results of post test inspection using conventional NDT and metallographic techniques are reported. (author)

  2. Dynamic analysis of the PEC fast reactor vessel: on-site tests and mathematical models

    International Nuclear Information System (INIS)

    Zola, Maurizio; Martelli, Alessandro; Maresca, Giuseppe; Masoni, Paolo; Scandola, Giani; Descleves, Pierre

    1988-01-01

    This paper presents the main features and results of the on-site dynamic tests and the related numerical analysis carried out for the PEC reactor vessel. The purpose is to provide an example of on-site testing of large components, stressing the problems encountered during the experiments, as well as in the processing phase of the test results and for the comparisons between calculations and measurements. Tests, performed by ISMES on behalf of ENEA, allowed the dynamic response of the empty vessel to be measured, thus providing data for the verification of the numerical models of the vessel supporting structure adopted in the PEC reactor-block seismic analysis. An axisymmetric model of the vessel, implemented in the vessel, implemented in the NOVAK code, had been developed in the framework of the detailed numerical studies performed by NOVATOME (again on behalf of ENEA), to check the beam schematization with fluid added mass model adopted by ANSALDO in SAP-IV and ANSYS for the reactor-block design calculations. Furthermore, a numerical model, describing vessel supporting structure in detail, was also developed by ANSALDO and implemented in the SAP-IV code. The test conditions were analysed by use of these and the design models. Comparisons between calculations and measurements showed particularly good agreement with regard to first natural frequency of the vessel and rocking stiffness of the vessel supporting structure, i.e. those parameters on which vessel seismic amplification mainly depends: this demonstrated the adequacy of the design analysis to correctly calculate the seismic motion at the PEC core diagrid. (author)

  3. Design Improvement of Double Pressure Vessel in the In-pile Test Section

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Heo, Sung-Ho; Joung, Chang-Young; Kim, Ka-Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    To carry out an irradiation test of nuclear fuels, a nuclear fuel test rig should be fabricated and installed in the in-pile test section (IPS), which is installed in the reactor hall. While carrying out an irradiation test, sealing out coolant which passes through the test rig is one of the most important issues. In particular, although the double pressure vessel is assembled with the IPS head by two o-rings and six bolts, 15.5 MPa of highly pressurized coolant leaks through the gap between the vessel and IPS head. Because the temperature of the coolant in the test loop is 300 .deg. C , and the pool of HANARO is 40 .deg. C, the double pressure vessel is necessary to insulate them. Therefore, a new design to prevent the leakage of coolant needs to be developed. In this study, EB welding technique is considered to assemble the double pressure vessel and the IPS head, and their mechanical design is modified to enable the welding process. In this study, an improved design for sealing out the coolant at the pressure boundary between the double pressure vessel and the IPS head has been developed. An EB weld is applied to seal out the pressure boundary, and its sealing performance is verified by NDE, a cross section test, and a hydraulic pressure test. From the verification test results, the improved design can be used in fabricating the IPS for a nuclear fuel irradiation test.

  4. The design, fabrication, and testing of WETF high-quality, long-term-storage, secondary containment vessels

    International Nuclear Information System (INIS)

    Fisher, Kane J.

    2000-01-01

    Los Alamos National Laboratory's Weapons Engineering Tritium Facility (WETF) requires secondary containment vessels to store primary tritium containment vessels. The primary containment vessel provides the first boundary for tritium containment. The primary containment vessel is stored within a secondary containment vessel that provides the secondary boundary for tritium containment. WETF requires high-quality, long-term-storage, secondary tritium containment vessels that fit within a Mound-designed calorimeter. In order to qualify the WETF high-quality, long-term-storage, secondary containment vessels for use at WETF, steps have been taken to ensure the appropriate design, adequate testing, quality in fabrication, and acceptable documentation

  5. Evaluation of HFIR vessel surveillance data and hydro-test conditions

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Nanstad, R.K.

    1994-01-01

    Surveillance specimens for the High Flux Isotope Reactor (HFIR) pressure vessel were removed and tested during 1993, after the vessel had accumulated 701,469 MWd of operation. The data agree well with HFIR surveillance data obtained in previous years. In conjunction with this effort, the vessel hydro-test conditions were reevaluated and found to be more than adequate. In view of this result, and because there are economic incentives for reducing the frequency of hydro testing, an analysis was performed to determine the minimum permissible frequency. The value obtained is substantially less than that presently specified. It was also determined that a somewhat lower cooling-tower-basin temperature is acceptable (improves operational flexibility). In 1986, after ∼20 years of reactor operation, it was discovered that the vessel embrittlement rate was substantially greater than expected. Possible reasons for the accelerated rate are reviewed in this report

  6. Pendulum support of the W7-X plasma vessel: Design, tests, manufacturing, assembly, critical aspects, status

    Energy Technology Data Exchange (ETDEWEB)

    Missal, B., E-mail: bernd.missal@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Leher, F.; Schiller, T. [MAN Diesel and Turbo SE, Werftstraße 17, 94469 Deggendorf (Germany); Friedrich, P. [Universität Rostock, FB Maschinenbau und Schiffstechnik, Albert-Einsteins-Straße 2, 18051 Rostock (Germany); Capriccioli, A. [ENEA Frascati, Fusion Technology Unit, Frascati (Italy)

    2014-10-15

    Highlights: • Plasma vessel support has to allow vertical adjustment and horizontal passive movement. • Planar sliding tables with PTFE do not fulfill all requirements. • Pendulums can fulfill all requirements. • Geometry and material of spherical bearings had to be optimized in calculations and tests. • Optimized pendulums were manufactured and assembled. - Abstract: The superconducting helical advanced stellarator Wendelstein 7-X (W7-X) is under construction at the Max-Planck-Institut für Plasmaphysik (IPP) in Greifswald, Germany. The three dimensional shape of plasma will be generated by 50 non-planar magnetic coils. The plasma vessel geometry follows exactly this three dimensional shape of plasma. To ensure the superconductivity of coils a cryo vacuum has to be generated. Therefore the coils and their support structure are enclosed within the outer vessel. Plasma vessel, coil structures and outer vessel have to be supported separately. This paper will describe the vertical supports of plasma vessel which have to fulfill two special requirements, vertical adjustability and horizontal mobility. These two tasks will be carried out by plasma vessel supports (PVS) with hydraulic cylinders, special sliding tables during assembly and pendulum supports during operating phase. The paper will give an overview of design, calculation, tests, fabrication, assembly, critical aspects and status of PVS.

  7. Dynamic analysis of the PEC fast reactor vessel: On-site tests and mathematical models

    International Nuclear Information System (INIS)

    Zola, M.; Martelli, A.; Masoni, P.; Scandola, G.

    1988-01-01

    This paper presents the main features and results of the on-site dynamic tests and the related numerical analyses carried out for the PEC reactor vessel. The purpose is to provide an example of on-site testing of large components, stressing the problems encountered during the experiments, as well as in the processing phase of the test results and for the comparisons between calculations and measurements. Tests, performed by ISMES on behalf of ENEA, allowed the dynamic response of the empty vessel to be measured, thus providing data for the verification of the numerical models of the vessel supporting structure adopted in the PEC reactor-block seismic analysis. An axisymmetric model of the vessel, implemented in the NOVAX code, had been developed in the framework of the detailed numerical studies performed by NOVATOME (again on behalf of ENEA), to check the beam schematization with fluid added mass model adopted by ANSALDO in SAP-IV and ANSYS for the reactor-block design calculations. Furthermore, a numerical model, describing vessel supporting structure in detail, was also developed by ANSALDO and implemented in the SAP-IV code. The test conditions were analysed by use of these and the design models. Comparisons between calculations and measurements showed particularly good agreement with regard to first natural frequency of the vessel and rocking stiffness of the vessel supporting structure, i.e. those parameters on which vessel seismic amplification mainly depends: this demonstrated the adequacy of the design analysis to correctly calculate the seismic motion at the PEC core diagrid. (author). 5 refs, 23 figs, 4 tabs

  8. German boiler and pressure vessel codes and standards: materials, manufacture, testing, equipment, erection and operation

    International Nuclear Information System (INIS)

    Steffen, H.P.

    1987-01-01

    The methods by which the safety objectives on the operation of steam boilers and pressure vessels in Germany can be reached are set out in Technical Rules which are compiled and established in technical committees. Typical applications are described in the Technical Rules. A chart shows how the laws, provisions and Technical Rules for the sections 'steam boiler plant' and 'pressure vessels' are interlinked. This chapter concentrates on legal aspects, materials, manufacture, testing, erection and operation of boilers and pressure vessels in Germany. (U.K.)

  9. Analyses and testing of model prestressed concrete reactor vessels with built-in planes of weakness

    International Nuclear Information System (INIS)

    Dawson, P.; Paton, A.A.; Fleischer, C.C.

    1990-01-01

    This paper describes the design, construction, analyses and testing of two small scale, single cavity prestressed concrete reactor vessel models, one without planes of weakness and one with planes of weakness immediately behind the cavity liner. This work was carried out to extend a previous study which had suggested the likely feasibility of constructing regions of prestressed concrete reactor vessels and biological shields, which become activated, using easily removable blocks, separated by a suitable membrane. The paper describes the results obtained and concludes that the planes of weakness concept could offer a means of facilitating the dismantling of activated regions of prestressed concrete reactor vessels, biological shields and similar types of structure. (author)

  10. Test of safety injection supply by diesel generator under reactor vessel closed condition

    International Nuclear Information System (INIS)

    Zhang Hao; Bi Fengchuan; Che Junxia; Zhang Jianwen; Yang Bo

    2014-01-01

    The paper studied that the test of diesel generator full load take-up under the condition of actual safety injection and reactor vessel closed in Ningde nuclear project unit l. It is proved that test result accorded with design criteria, meanwhile, the test was removed from the key path of project schedule, which cut a huge cost. (authors)

  11. Local gamma ray events as tests of the antimatter theory of gamma ray bursts

    International Nuclear Information System (INIS)

    Sofia, S.; Wilson, R.E.

    1976-01-01

    Nearby examples of the antimatter 'chunks' postulated by Sofia and Van Horn to explain the cosmic gamma ray bursts may produce detectable gamma ray events when struck by solar system meteoroids. These events would have a much shorter time scale and higher energy spectrum than the bursts already observed. In order to have a reasonably high event rate, the local meteoroid population must extend to a distance from the Sun of the order of 0.1 pc, but the required distance could become much lower if the instrumental threshold is improved. The expected gamma ray flux for interaction of the antimatter bodies with the solar wind is also examined, and found to be far below present instrumental capabilities. (Auth.)

  12. Reliability of COPVs Accounting for Margin of Safety on Design Burst

    Science.gov (United States)

    Murthy, Pappu L.N.

    2012-01-01

    In this paper, the stress rupture reliability of Carbon/Epoxy Composite Overwrapped Pressure Vessels (COPVs) is examined utilizing the classic Phoenix model and accounting for the differences between the design and the actual burst pressure, and the liner contribution effects. Stress rupture life primarily depends upon the fiber stress ratio which is defined as the ratio of stress in fibers at the maximum expected operating pressure to actual delivered fiber strength. The actual delivered fiber strength is calculated using the actual burst pressures of vessels established through burst tests. However, during the design phase the actual burst pressure is generally not known and to estimate the reliability of the vessels calculations are usually performed based upon the design burst pressure only. Since the design burst is lower than the actual burst, this process yields a much higher value for the stress ratio and consequently a conservative estimate for the reliability. Other complications arise due to the fact that the actual burst pressure and the liner contributions have inherent variability and therefore must be treated as random variables in order to compute the stress rupture reliability. Furthermore, the model parameters, which have to be established based on stress rupture tests of subscale vessels or coupons, have significant variability as well due to limited available data and hence must be properly accounted for. In this work an assessment of reliability of COPVs including both parameter uncertainties and physical variability inherent in liner and overwrap material behavior is made and estimates are provided in terms of degree of uncertainty in the actual burst pressure and the liner load sharing.

  13. Long- and short-term trends in vessel conditioning of TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    LaMarche, P.H.; Dylla, H.F.; Bell, M.G.

    1986-10-01

    We have investigated trends in the conditioning of the Tokamak Fusion Test Reactor (TFTR) vacuum vessel during the May 1984 to April 1985 run period. The initial conditioning of the vessel, consisting of glow discharge cleaning (GDC) and pulse discharge cleaning (PDC) in concert with a 150 0 C vessel bakeout, is necessary to assure plasma operation after atmospheric venting. A long-term conditioning process, ascribed to limiter conditioning, effectively improves operational conditions during the course of the run. Over several thousand high power plasma discharges, the improvement was documented by using standard parameter (fiducial) plasma discharges. Several techniques demonstrated short-term improvements in vessel conditioning during this time period, including: Cr gettering and programming the plasma position relative to the limiter contact area

  14. Integrity assessment of TAPS reactor pressure vessel at extended EOL using surveillance test results

    International Nuclear Information System (INIS)

    Chatterjee, S.; Shah, Priti Kotak

    2008-05-01

    Integrity assessment of pressure vessels of nuclear reactors (RPV) primarily concentrates on the prevention of brittle failure and conditions are defined under which brittle failure can be excluded. Accordingly, two approaches based on Transition Temperature Concept and Fracture Mechanics Concept were adopted using the impact test results of three credible surveillance data sets obtained from the surveillance specimens of Tarapur Atomic Power Station. RT NDT data towards end of life (EOL) were estimated from the impact test results in accordance with the procedures of USNRC Regulatory Guide 1.99, Rev. 2 and were used as primary input for assessment of the vessel integrity. SA302B (nickel modified) steel cladded with stainless steel is used as the pressure vessel material for the two 210 MWe boiling water reactors of the Tarapur Atomic Power Station (TAPS). The reactors were commissioned during the year 1969. The chemical compositions of SA302B (modified) steel used in fabricating the vessel and the specified tensile property and the Charpy impact property requirements of the steel broadly meet ASME specified requirements. Therefore, the pressure temperature limit curves prescribed by General Electric (G.E.) were compared with those as obtained using procedures of ASME Section XII, Appendix G. The tensile and the Charpy impact properties at 60 EFPY of vessel operation as derived from the surveillance specimens even fulfilled the specified requirements for the virgin material of ASME. Integrity assessment carried out using the two approaches indicated the safety of the vessel for continued operation up to 60 EFPY. (author)

  15. Xylem resistance to embolism: presenting a simple diagnostic test for the open vessel artefact.

    Science.gov (United States)

    Torres-Ruiz, José M; Cochard, Hervé; Choat, Brendan; Jansen, Steven; López, Rosana; Tomášková, Ivana; Padilla-Díaz, Carmen M; Badel, Eric; Burlett, Regis; King, Andrew; Lenoir, Nicolas; Martin-StPaul, Nicolas K; Delzon, Sylvain

    2017-07-01

    Xylem vulnerability to embolism represents an essential trait for the evaluation of the impact of hydraulics in plant function and ecology. The standard centrifuge technique is widely used for the construction of vulnerability curves, although its accuracy when applied to species with long vessels remains under debate. We developed a simple diagnostic test to determine whether the open-vessel artefact influences centrifuge estimates of embolism resistance. Xylem samples from three species with differing vessel lengths were exposed to less negative xylem pressures via centrifugation than the minimum pressure the sample had previously experienced. Additional calibration was obtained from non-invasive measurement of embolism on intact olive plants by X-ray microtomography. Results showed artefactual decreases in hydraulic conductance (k) for samples with open vessels when exposed to a less negative xylem pressure than the minimum pressure they had previously experienced. X-Ray microtomography indicated that most of the embolism formation in olive occurs at xylem pressures below -4.0 MPa, reaching 50% loss of hydraulic conductivity at -5.3 MPa. The artefactual reductions in k induced by centrifugation underestimate embolism resistance data of species with long vessels. A simple test is suggested to avoid this open vessel artefact and to ensure the reliability of this technique in future studies. © 2017 The Authors. New Phytologist © 2017 New Phytologist Trust.

  16. Design of vacuum vessel for Indian Test Facility (INTF) for 100 keV neutral beams

    International Nuclear Information System (INIS)

    Joshi, Jaydeep; Yadav, Ashish; Gangadharan, Roopesh; Prasad, Rambilas; Ulahannan, Shino; Rotti, Chandramouli; Bandyopadhyay, Mainak; Chakraborty, Arun

    2015-01-01

    Highlights: • Thickness calculation and optimization for the main shell, ducts, Dishends and top lid on the main shell. • Nozzle and flange design for the port openings. • Support structure design for the main shell and ducts. • FEA validation of the INTF vessel for operational, seismic and lifting condition. - Abstract: The Indian Test Facility (INTF) vacuum vessel is designed to install a full-scale test set-up of Diagnostic Neutral Beam (DNB) [1] for the qualification of beam parameters and the behavior of beam-line components prior to installation and operation in ITER. Vacuum vessel is designed in cylindrical shape having length of ∼9 m with diameter of ∼4.5 m and has a detachable top-lid for mounting as well as removal of internal components during installation and maintenance phases. The Vessel has hemispherical dish-ends with large openings for high-voltage bushing on one side and duct on another side. Vessel is provided with openings for hydraulic, cryo, gas-feed and diagnostics. Vessel duct is composed of three segments with length ranges from 3 m to 5 m with diameter of ∼1.5 m and one vessel at the end to house the second calorimeter. The objective of this paper is to present the design and analysis of vacuum vessel, with respect to its functional and operational requirements. The design calculations are done as per ASME-BPVC SectionVIII-Div.1 and subsequently Finite Element Analysis (FEM) method has been adopted to verify the design.

  17. Design of vacuum vessel for Indian Test Facility (INTF) for 100 keV neutral beams

    Energy Technology Data Exchange (ETDEWEB)

    Joshi, Jaydeep, E-mail: Jaydeep.joshi@iter-india.org [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India); Yadav, Ashish; Gangadharan, Roopesh [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India); Prasad, Rambilas [Madan Mohan Malaviya University of Technology, Gorakhpur, Uttar Pradesh 273001 (India); Ulahannan, Shino [Airframe Aerodesigns Pvt. Ltd., HAL Airport Exit Road, Old Airport Road, Bengaluru 17 (India); Rotti, Chandramouli; Bandyopadhyay, Mainak; Chakraborty, Arun [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India)

    2015-10-15

    Highlights: • Thickness calculation and optimization for the main shell, ducts, Dishends and top lid on the main shell. • Nozzle and flange design for the port openings. • Support structure design for the main shell and ducts. • FEA validation of the INTF vessel for operational, seismic and lifting condition. - Abstract: The Indian Test Facility (INTF) vacuum vessel is designed to install a full-scale test set-up of Diagnostic Neutral Beam (DNB) [1] for the qualification of beam parameters and the behavior of beam-line components prior to installation and operation in ITER. Vacuum vessel is designed in cylindrical shape having length of ∼9 m with diameter of ∼4.5 m and has a detachable top-lid for mounting as well as removal of internal components during installation and maintenance phases. The Vessel has hemispherical dish-ends with large openings for high-voltage bushing on one side and duct on another side. Vessel is provided with openings for hydraulic, cryo, gas-feed and diagnostics. Vessel duct is composed of three segments with length ranges from 3 m to 5 m with diameter of ∼1.5 m and one vessel at the end to house the second calorimeter. The objective of this paper is to present the design and analysis of vacuum vessel, with respect to its functional and operational requirements. The design calculations are done as per ASME-BPVC SectionVIII-Div.1 and subsequently Finite Element Analysis (FEM) method has been adopted to verify the design.

  18. Design and performance tests of gas circulation heating of JT-60U vacuum vessel

    International Nuclear Information System (INIS)

    Yotsuga, M.; Masuzaki, T.; Sago, H.; Nishikane, M.; Uchikawa, T.; Iritani, Y.; Murakami, T.; Horiike, H.; Neyatani, Y.; Ninomiya, H.; Matsukawa, M.; Ando, T.; Miyachi, I.

    1992-01-01

    This paper reports that in the final stage of construction of the upgraded JT-60 device (JT-60U), baking tests of the vacuum vessel was performed. The vessel torus was heated-up to 300 degrees C by means of the nitrogen gas circulation system and electric heaters mounted on the outboard solid wall of the vessel. The design of the gas flow channels inside the double-wall structure of the vessel was done based on flow model tests, fluid analysis, and flow network analysis. The results of the baking tests were satisfactory. In maintaining 300 degrees C bake-out temperature, required heating power of the gas circulation system and outboard heaters was 520kW and 50kW, respectively. The temperature distribution over the vessel wall was within 300 ± 30 degrees C. It was also shown or suggested that heat-up and cool-down time is about 30 hours. The baking tests data have been reflected on operations for plasma experiments

  19. Mechanical behaviour of the reactor vessel support of a pressurized water reactor: tests and analysis

    International Nuclear Information System (INIS)

    Bolvin, M.; L'huby, Y.; Quillico, J.J.; Humbert, J.M.; Thomas, J.P.; Hugenschmitt, R.

    1985-08-01

    The PWR reactor vessel is supported by a steel ring laying on the reactor pit. This support has to ensure a good behaviour of the vessel in the event of accidental conditions (earthquake and pipe rupture). A new evolution of the evaluation methods of the applied forces has shown a significant increase in the design loads used until now. In order to take into account these new forces, we carried out a test on a representative mock-up of the vessel support (scale 1/6). This test was performed by CEA, EDF and FRAMATOME. Several static equivalent forces were applied on the experimental mock-up. Displacements and strains were simultaneously recorded. The results of the test have enabled to justify the design of the pit and the ring, to show up a wide safety margin until the collapse of the structures and to check our hypothesis about the transmission of the forces between the ring and the pit

  20. Properties important to mixing and simulant recommendations for WTP full-scale vessel testing

    Energy Technology Data Exchange (ETDEWEB)

    Poirier, M. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Martino, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-12-01

    Full Scale Vessel Testing (FSVT) is being planned by Bechtel National, Inc., to demonstrate the ability of the standard high solids vessel design (SHSVD) to meet mixing requirements over the range of fluid properties planned for processing in the Pretreatment Facility (PTF) of the Hanford Waste Treatment and Immobilization Plant (WTP). Testing will use simulated waste rather than actual Hanford waste. Therefore, the use of suitable simulants is critical to achieving the goals of the test program. WTP personnel requested the Savannah River National Laboratory (SRNL) to assist with development of simulants for use in FSVT. Among the tasks assigned to SRNL was to develop a list of waste properties that are important to pulse-jet mixer (PJM) performance in WTP vessels with elevated concentrations of solids.

  1. Virtual simulation of maneuvering captive tests for a surface vessel

    Directory of Open Access Journals (Sweden)

    Ahmad Hajivand

    2015-09-01

    Full Text Available Hydrodynamic derivatives or coefficients are required to predict the maneuvering characteristics of a marine vehicle. These derivatives are obtained numerically for a DTMB 5512 model ship by virtual simulating of captive model tests in a CFD environment. The computed coefficients are applied to predict the turning circle and zigzag maneuvers of the model ship. The comparison of the simulated results with the available experimental data shows a very good agreement among them. The simulations show that the CFD is precise and affordable tool at the preliminary design stage to obtain maneuverability performance of a marine vehicles.

  2. Experimental system description for air-water CCFL tests of the 161-rod FLECHT-SEASET test vessel upper plenum

    International Nuclear Information System (INIS)

    Fogdall, S.P.; Anderson, J.L.

    1983-01-01

    A series of countercurrent flow limiting (CCFL) experiments has been performed by EG and G Idaho, Inc. in the Steam-Air-Water (SAW) test facility at the Idaho National Engineering Laboratory on behalf of the US Nuclear Regulatory Commission (NRC). Tests were performed in a mockup of the vessel for the 161-Rod Systems Effects Test (SET) facility of the FLECHT-SEASET program, conducted by the Westinghouse Electric Corporation. Westinghouse and the NRC will use the test results to provide a CCFL correlation to predict the flooding behavior in the upper plenum of the SET vessel. This paper presents a description of the experimental system and the test conduct, including data validation and uncertainty analysis. The test objectives centered on experimentally obtaining coefficients in the Wallis correlation for flooding with the specific vessel geometry. The test conditions and vessel configuration are described and the design of the test loop, instrumentation, and data acquisition are discussed. The establishment of a test point and the resultant data are described

  3. Hydraulic burst tests at elevated temperatures on Zircaloy cladding from fuel rods irradiated in the Winfrith SGHWR

    International Nuclear Information System (INIS)

    Garlick, A.; Hindmarch, P.

    1980-09-01

    Closed-end hydraulic burst tests have been carried out at 613K on lengths of cladding cut from fuel rods that had been irradiated in the SGHWR to 25 n/m 2 . The effects of reactor exposure on the mechanical properties of the Zircaloy cladding, initially in the stress-relieved and fully recrystallised conditions, have been evaluated from measurements of the 0.2% proof stress, the ultimate burst stress, the total circumferential elongation and the reduction in wall thickness at fracture. It is shown that after irradiation, the measured strength properties of stress-relieved cladding remained higher than for that in the fully recrystallised condition, although the large differences observed before irradiation were considerably reduced. The irradiation-induced increase in proof stress measured during these tests was compared with US results from uniaxial tensile tests and, after correcting for the effect of stress-ratio, it is concluded that close agreement exists between the two sets of data for Zircaloy in the fully recrystallised condition. In contrast, the agreement for stress-relieved Zircaloy is less good, although the maximum increase in proof stress after high neutron doses for this material is similar for data from the two sources. After irradiation, the ductility of fully recrystallised Zircaloy remained higher than that of stress-relieved material and there was no evidence to suggest that a serious loss of ductility had occurred for Zircaloy in either condition of heat-treatment as a result of reactor exposure. (author)

  4. Mock-up tests of rail-mounted vehicle type in-vessel transporter/manipulator

    International Nuclear Information System (INIS)

    Oka, K.; Kakaudate, S.; Fukatsu, S.

    1995-01-01

    A rail-mounted vehicle system has been developed for remote maintenance of in-vessel components for fusion experimental reactor. In this system, a rail deploying/storing system is installed at outside of the reactor core and used to deploy a rail transporter and vehicle/manipulator for the in-vessel maintenance. A prototype of the rail deploying/storing system has been fabricated for mockup tests. This paper describes structural design of the prototypical rail deploying/storing system and results of the performance tests such as payload capacity, position control and rail deployment/storage performance

  5. Nuclear reactor pressure vessel surveillance capsule examinations. Application of American Society for Testing and Materials Standards

    International Nuclear Information System (INIS)

    Perrin, J.S.

    1978-01-01

    A series of pressure vessel surveillance capsules is installed in each commercial nuclear power plant in the United States. A capsule typically contains neutron dose meters, thermal monitors, tensile specimens, and Charpy V-notch impact specimens. In order to determine property changes of the pressure vessel resulting from irradiation, surveillance capsules are periodically removed during the life of a reactor and examined. There are numerous standards, regulations, and codes governing US pressure vessel surveillance capsule programmes. These are put out by the US Nuclear Regulatory Commission, the Boiler and Pressure Vessel Committee of the American Society of Mechanical Engineers, and the American Society for Testing and Materials (ASTM). A majority of the pertinent ASTM standards are under the jurisdiction of ASTM Committee E-10 on Nuclear Applications and Measurements of Radiation Effects. The standards, regulations, and codes pertaining to pressure vessel surveillance play an important role in ensuring reliability of the nuclear pressure vessels. ASTM E 185-73 is the Standard Recommended Practice for Surveillance Tests for Nuclear Reactors. This standard recommends procedures for both the irradiation and subsequent testing of surveillance capsules. ASTM E 185-73 references many additional specialized ASTM standards to be followed in specific areas of a surveillance capsule examination. A key element of surveillance capsule programmes is the Charpy V-notch impact test, used to define curves of fracture behaviour over a range of temperatures. The data from these tests are used to define the adjusted reference temperature used in determining pressure-temperature operating curves for a nuclear power plant. (author)

  6. Predicting crack instability behavior of burst tests from small specimens for irradiated Zr-2.5Nb pressure tubes

    International Nuclear Information System (INIS)

    Davies, P.H.

    1997-01-01

    A scaling approach, based on the deformation J-integral at maximum load obtained from small specimens, is proposed for predicting the crack instability behavior of burst tests on irradiated Zr-2.5Nb pressure tubes. An assessment of this approach is carried out by comparison with other toughness criteria such as the modified J-integral and the plastic work dissipation rate approach. The largest discrepancy between the different parameters occurs for materials of intermediate toughness which exhibit the most stable crack growth and tunnelling up to maximum load. A study of one material of intermediate toughness suggests crack-front tunnelling has a significant influence on the results obtained from the 17-mm-wide specimens. It is shown that for a tube of intermediate toughness the different approaches can significantly underpredict the extent of stable crack growth before instability in a burst test even after correcting for tunnelling. The usefulness of a scaling approach in reducing the discrepancy between the small- and large-scale specimen results for this material is demonstrated

  7. Scaled Testing to Evaluate Pulse Jet Mixer Performance in Waste Treatment Plant Mixing Vessels

    International Nuclear Information System (INIS)

    Fort, James A.; Meyer, Perry A.; Bamberger, Judith A.; Enderlin, Carl W.; Scott, Paul A.; Minette, Michael J.; Gauglitz, Phillip A.

    2010-01-01

    The Waste Treatment and Immobilization Plant (WTP) at Hanford is being designed and built to pre-treat and vitrify the waste in Hanford's 177 underground waste storage tanks. Numerous process vessels will hold waste at various stages in the WTP. These vessels have pulse jet mixer (PJM) systems. A test program was developed to evaluate the adequacy of mixing system designs in the solids-containing vessels in the WTP. The program focused mainly on non-cohesive solids behavior. Specifically, the program addressed the effectiveness of the mixing systems to suspend settled solids off the vessel bottom, and distribute the solids vertically. Experiments were conducted at three scales using various particulate simulants. A range of solids loadings and operational parameters were evaluated, including jet velocity, pulse volume, and duty cycle. In place of actual PJMs, the tests used direct injection from tubes with suction at the top of the tank fluid. This gave better control over the discharge duration and duty cycle and simplified the facility requirements. The mixing system configurations represented in testing varied from 4 to 12 PJMs with various jet nozzle sizes. In this way the results collected could be applied to the broad range of WTP vessels with varying geometrical configurations and planned operating conditions. Data for 'just-suspended velocity', solids cloud height, and solids concentration vertical profile were collected, analyzed, and correlated. The correlations were successfully benchmarked against previous large-scale test results, then applied to the WTP vessels using reasonable assumptions of anticipated waste properties to evaluate adequacy of the existing mixing system designs.

  8. Ultimate load design and testing of a cylindrical prestressed concrete vessel

    International Nuclear Information System (INIS)

    Stefanou, G.D.

    1982-01-01

    The object of this research was to design, construct and test to failure a prestressed concrete pressure vessel model that could be used to investigate the behavior of a full scale structure underworking and ultimate load. The properties and the design of the model was based generally on full scale vessels already constructed to house the nuclear reactors used in atomic power stations. To design the model the ultimate load approach was adopted throughout. All load factors associated with the prestressing have been defined and kept to a minimum in order that the vessel's behavior may be predicted. The tests on the vessel were carried out first on the elastic range to observe its behavior at working load and then at the ultimate range to observe the modes of failure and compare the actual results in both cases with the predicted values. Although full agreement between observed results and predicted values was not obtained, the conclusions drawn from the study were useful for the design of full scale vessels. (author)

  9. A fracture mechanics and reliability based method to assess non-destructive testings for pressure vessels

    International Nuclear Information System (INIS)

    Kitagawa, Hideo; Hisada, Toshiaki

    1979-01-01

    Quantitative evaluation has not been made on the effects of carrying out preservice and in-service nondestructive tests for securing the soundness, safety and maintainability of pressure vessels, spending large expenses and labor. Especially the problems concerning the time and interval of in-service inspections lack the reasonable, quantitative evaluation method. In this paper, the problems of pressure vessels are treated by having developed the analysis method based on reliability technology and probability theory. The growth of surface cracks in pressure vessels was estimated, using the results of previous studies. The effects of nondestructive inspection on the defects in pressure vessels were evaluated, and the influences of many factors, such as plate thickness, stress, the accuracy of inspection and so on, on the effects of inspection, and the method of evaluating the inspections at unequal intervals were investigated. The analysis of reliability taking in-service inspection into consideration, the evaluation of in-service inspection and other affecting factors through the typical examples of analysis, and the review concerning the time of inspection are described. The method of analyzing the reliability of pressure vessels, considering the growth of defects and preservice and in-service nondestructive tests, was able to be systematized so as to be practically usable. (Kako, I.)

  10. Testing plan for critical heat flux measurement during in-vessel retention

    International Nuclear Information System (INIS)

    Aoki, Kazuyoshi; Iwaki, Chikako; Sato, Hisaki; Mimura, Satoshi; Kanamori, Daisuke

    2015-01-01

    In-Vessel Retention (IVR) is a method to maintain molten debris in a reactor vessel (RV) by RV outer surface cooling. Structural integrity of RV and cooling capacity on RV outer surface are important to verify IVR strategy. Critical Heat Flux (CHF) data is necessary to estimate cooling capacity on the RV outer surface. And there are some CHF data to estimate cooling capacity on the RV outer surface. However, these data were obtained for specific plants. Thus, the objective of this study is developing a CHF correlation for various PWR plants. The objectives of this paper are developing test equipment and testing plan for the CHF correlation. Firstly, plant conditions during severe accidents were organized. Then, ranges of testing parameters were estimated with the plant conditions. And specifications of the test equipment were set to cover the range of parameters. Secondly, testing cases were set based on design of experiments. The test cases are suitable to develop experimental correlations. (author)

  11. Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required by Practices E185 and E2215 for the surveillance of nuclear reactor vessels. The supplemental mechanical property tests outlined permit the acquisition of additional information on radiation-induced changes in fracture toughness, notch ductility, and yield strength properties of the reactor vessel steels. 1.2 This guide provides recommendations for the preparation of test specimens for irradiation, and identifies special precautions and requirements for reactor surveillance operations and postirradiation test planning. Guidance on data reduction and computational procedures is also given. Reference is made to other ASTM test methods for the physical conduct of specimen tests and for raw data acquisition.

  12. Verification of the analytical fracture assessments methods by a large scale pressure vessel test

    Energy Technology Data Exchange (ETDEWEB)

    Keinanen, H; Oberg, T; Rintamaa, R; Wallin, K

    1988-12-31

    This document deals with the use of fracture mechanics for the assessment of reactor pressure vessel. Tests have been carried out to verify the analytical fracture assessment methods. The analysis is focused on flaw dimensions and the scatter band of material characteristics. Results are provided and are compared to experimental ones. (TEC).

  13. Mechanical properties of reactor pressure vessel steels studied by static and dynamic torsion tests

    International Nuclear Information System (INIS)

    Munier, A.; Maamouri, M.; Schaller, R.; Mercier, O.

    1993-01-01

    Internal friction measurements and torsional plastic deformation tests have been performed in reactor pressure vessel steels (unirradiated, irradiated and irradiated/annealed specimens). The results of these experiments have been interpreted with help of transmission electron microscopy observations (conventional and in situ). It is shown how the interactions between screw dislocations and obstacles (Peierls valleys, impurities and precipitates) could explain the low temperature hardening and the irradiation embrittlement of ferritic steels. In addition, it appears that the nondestructive internal friction technique could be used advantageously to follow the evolution of the material properties under irradiation, as for instance the irradiation embrittlement of the reactor pressure vessel steels. (orig.)

  14. Manufacturing, assembly and tests of SPIDER Vacuum Vessel to develop and test a prototype of ITER neutral beam ion source

    Energy Technology Data Exchange (ETDEWEB)

    Zaccaria, Pierluigi, E-mail: pierluigi.zaccaria@igi.cnr.it [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete S.p.A.), Padova (Italy); Valente, Matteo; Rigato, Wladi; Dal Bello, Samuele; Marcuzzi, Diego; Agostini, Fabio Degli; Rossetto, Federico; Tollin, Marco [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete S.p.A.), Padova (Italy); Masiello, Antonio [Fusion for Energy F4E, Barcelona (Spain); Corniani, Giorgio; Badalocchi, Matteo; Bettero, Riccardo; Rizzetto, Dario [Ettore Zanon S.p.A., Schio (VI) (Italy)

    2015-10-15

    Highlights: • The SPIDER experiment aims to qualify and optimize the ion source for ITER injectors. • The large SPIDER Vacuum Vessel was built and it is under testing at the supplier. • The main working and assembly steps for production are presented in the paper. - Abstract: The SPIDER experiment (Source for the Production of Ions of Deuterium Extracted from an RF plasma) aims to qualify and optimize the full size prototype of the negative ion source foreseen for MITICA (full size ITER injector prototype) and the ITER Heating and Current Drive Injectors. Both SPIDER and MITICA experiments are presently under construction at Consorzio RFX in Padova (I), with the financial support from IO (ITER Organization), Fusion for Energy, Italian research institutions and contributions from Japan and India Domestic Agencies. The vacuum vessel hosting the SPIDER in-vessel components (Beam Source and calorimeters) has been manufactured, assembled and tested during the last two years 2013–2014. The cylindrical vessel, about 6 m long and 4 m in diameter, is composed of two cylindrical modules and two torispherical lids at the ends. All the parts are made by AISI 304 L stainless steel. The possibility of opening/closing the vessel for monitoring, maintenance or modifications of internal components is guaranteed by bolted junctions and suitable movable support structures running on rails fixed to the building floor. A large number of ports, about one hundred, are present on the vessel walls for diagnostic and service purposes. The main working steps for construction and specific technological issues encountered and solved for production are presented in the paper. Assembly sequences and tests on site are furthermore described in detail, highlighting all the criteria and requirements for correct positioning and testing of performances.

  15. Manufacturing, assembly and tests of SPIDER Vacuum Vessel to develop and test a prototype of ITER neutral beam ion source

    International Nuclear Information System (INIS)

    Zaccaria, Pierluigi; Valente, Matteo; Rigato, Wladi; Dal Bello, Samuele; Marcuzzi, Diego; Agostini, Fabio Degli; Rossetto, Federico; Tollin, Marco; Masiello, Antonio; Corniani, Giorgio; Badalocchi, Matteo; Bettero, Riccardo; Rizzetto, Dario

    2015-01-01

    Highlights: • The SPIDER experiment aims to qualify and optimize the ion source for ITER injectors. • The large SPIDER Vacuum Vessel was built and it is under testing at the supplier. • The main working and assembly steps for production are presented in the paper. - Abstract: The SPIDER experiment (Source for the Production of Ions of Deuterium Extracted from an RF plasma) aims to qualify and optimize the full size prototype of the negative ion source foreseen for MITICA (full size ITER injector prototype) and the ITER Heating and Current Drive Injectors. Both SPIDER and MITICA experiments are presently under construction at Consorzio RFX in Padova (I), with the financial support from IO (ITER Organization), Fusion for Energy, Italian research institutions and contributions from Japan and India Domestic Agencies. The vacuum vessel hosting the SPIDER in-vessel components (Beam Source and calorimeters) has been manufactured, assembled and tested during the last two years 2013–2014. The cylindrical vessel, about 6 m long and 4 m in diameter, is composed of two cylindrical modules and two torispherical lids at the ends. All the parts are made by AISI 304 L stainless steel. The possibility of opening/closing the vessel for monitoring, maintenance or modifications of internal components is guaranteed by bolted junctions and suitable movable support structures running on rails fixed to the building floor. A large number of ports, about one hundred, are present on the vessel walls for diagnostic and service purposes. The main working steps for construction and specific technological issues encountered and solved for production are presented in the paper. Assembly sequences and tests on site are furthermore described in detail, highlighting all the criteria and requirements for correct positioning and testing of performances.

  16. Reactor vessel dismantling at the high flux materials testing reactor Petten

    International Nuclear Information System (INIS)

    Tas, A.; Teunissen, G.

    1986-01-01

    The project of replacing the reactor vessel of the high flux materials testing reactor (HFR) originated in 1974 when results of several research programs confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report describes the dismantling philosophy and organisation, the design of special underwater equipment, the dismantling of the reactor vessel and thermal column, and the conditioning and shielding activities resulting in a working area for the installation of the new vessel with no access limitations due to radiation. Finally an overview of the segmentation, waste disposal and radiation exposure is given. The total dismantling, segmentation and conditioning activities resulted in a total collective radiation dose of 300 mSv. (orig.) [de

  17. Fabrication of High Temperature and High Pressure Vessel for the Fuel Test

    International Nuclear Information System (INIS)

    Park, Kook Nam; Lee, Jong Min; Sim, Bong Shick; Shon, Jae Min; Ahn, Seung Ho; Yoo, Seong Yeon

    2007-01-01

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR and CANDU nuclear power plants has been developed and installed in HANARO, KAERI. It is consisted of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS which is located inside the pool is divided into 3-parts; they are in-pool pipes, IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The localization of the IVA is achieved by manufacturing through local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique of the instrument lines has been checked for its functionality and yield. A IVA has been manufactured by local technique and will be finally tested under out of the high temperature and high pressure test

  18. Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program. PSF Blind Test workshop minutes. Summary

    International Nuclear Information System (INIS)

    Guthrie, G.L.; Lippincott, E.P.; McGarry, E.D.

    1984-01-01

    A ''Blind Test'' workshop was held on April 9-11, 1984, at the Holiday Inn in Richland, WA. At the workshop, participant groups compared ''Blind'' calculations with existing data which was unavailable to them at the time the calculations were made. The purpose of the exercise was to allow each participant group to test the group's ability to predict ''in-wall'' mechanical property degradation for a simulated nuclear reactor pressure vessel irradiation

  19. TESTING MODELS FOR THE SHALLOW DECAY PHASE OF GAMMA-RAY BURST AFTERGLOWS WITH POLARIZATION OBSERVATIONS

    Energy Technology Data Exchange (ETDEWEB)

    Lan, Mi-Xiang; Dai, Zi-Gao [School of Astronomy and Space Science, Nanjing University, Nanjing 210093 (China); Wu, Xue-Feng, E-mail: dzg@nju.edu.cn [Purple Mountain Observatory, Chinese Academy of Sciences, Nanjing 210008 (China)

    2016-08-01

    The X-ray afterglows of almost one-half of gamma-ray bursts have been discovered by the Swift satellite to have a shallow decay phase of which the origin remains mysterious. Two main models have been proposed to explain this phase: relativistic wind bubbles (RWBs) and structured ejecta, which could originate from millisecond magnetars and rapidly rotating black holes, respectively. Based on these models, we investigate polarization evolution in the shallow decay phase of X-ray and optical afterglows. We find that in the RWB model, a significant bump of the polarization degree evolution curve appears during the shallow decay phase of both optical and X-ray afterglows, while the polarization position angle abruptly changes its direction by 90°. In the structured ejecta model, however, the polarization degree does not evolve significantly during the shallow decay phase of afterglows whether the magnetic field configuration in the ejecta is random or globally large-scale. Therefore, we conclude that these two models for the shallow decay phase and relevant central engines would be testable with future polarization observations.

  20. Gravity wave and neutrino bursts from stellar collapse: A sensitive test of neutrino masses

    International Nuclear Information System (INIS)

    Arnaud, N.; Barsuglia, M.; Bizouard, M.A.; Cavalier, F.; Davier, M.; Hello, P.; Pradier, T.

    2002-01-01

    New methods are proposed with the goal to determine absolute neutrino masses from the simultaneous observation of the bursts of neutrinos and gravitational waves emitted during a stellar collapse. It is shown that the neutronization electron neutrino flash and the maximum amplitude of the gravitational wave signal are tightly synchronized with the bounce occurring at the end of the core collapse on a time scale better than 1 ms. The existing underground neutrino detectors (SuperKamiokande, SNO,...) and the gravity wave antennas soon to operate (LIGO, VIRGO,...) are well matched in their performance for detecting galactic supernovae and for making use of the proposed approach. Several methods are described, which apply to the different scenarios depending on neutrino mixing. Given the present knowledge on neutrino oscillations, the methods proposed are sensitive to a mass range where neutrinos would essentially be mass degenerate. The 95% C.L. upper limit which can be achieved varies from 0.75 eV/c 2 for large ν e survival probabilities to 1.1 eV/c 2 when in practice all ν e 's convert into ν μ 's or ν τ 's. The sensitivity is nearly independent of the supernova distance

  1. Experimental tests on buckling of ellipsoidal vessel heads subjected to internal pressure

    International Nuclear Information System (INIS)

    Roche, R.L.; Alix, M.

    1980-05-01

    Tests were performed on 17 ellipsoidal vessel heads of three different materials and different geometries. The results include the following: 1) Accurate definition of the geometry and particularly a direct measurement of the thickness along the meridian. 2) The properties of the material of each head, obtained from test specimens cut from the head itself after the test. 3) The recording of deflection/pressure curves with indication of the pressure at which buckling occurred. These results can be used for validation and qualification of methods for calculating the buckling load when plasticity occurs before buckling. It was possible to develop an empirical equation representing the experimental results obtained with satisfactory accuracy. This equation may be useful in pressure vessel design

  2. Mock-up test of remote controlled dismantling apparatus for large-sized vessels (contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Myodo, Masato; Miyajima, Kazutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Okane, Shogo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-03-01

    The Remote dismantling apparatus, which is equipped with multi-units for functioning of washing, cutting, collection of cut pieces and so on, has been constructed to dismantle the large-sized vessels in the JAERI's Reprocessing Test Facility (JRTF). The apparatus has five-axis movement capability and its operation is performed remotely. The mock-up tests were performed to evaluate the applicability of the apparatus to actual dismantling activities by using the mock-ups of LV-3 and LV-5 in the facility. It was confirmed that each unit was satisfactory functioned by remote operation. Efficient procedures for dismantling the large-sized vessel was studied and various date was obtained in the mock-up tests. This apparatus was found to be applicable for the actual dismantling activity in JRTF. (author)

  3. Mock-up test of remote controlled dismantling apparatus for large-sized vessels (contract research)

    International Nuclear Information System (INIS)

    Myodo, Masato; Miyajima, Kazutoshi; Okane, Shogo

    2001-03-01

    The Remote dismantling apparatus, which is equipped with multi-units for functioning of washing, cutting, collection of cut pieces and so on, has been constructed to dismantle the large-sized vessels in the JAERI's Reprocessing Test Facility (JRTF). The apparatus has five-axis movement capability and its operation is performed remotely. The mock-up tests were performed to evaluate the applicability of the apparatus to actual dismantling activities by using the mock-ups of LV-3 and LV-5 in the facility. It was confirmed that each unit was satisfactory functioned by remote operation. Efficient procedures for dismantling the large-sized vessel was studied and various date was obtained in the mock-up tests. This apparatus was found to be applicable for the actual dismantling activity in JRTF. (author)

  4. The 5th surveillance testing for Kori unit 1 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)

  5. Babcock experience of automated ultrasonic non-destructive testing of PWR pressure vessels during manufacture

    International Nuclear Information System (INIS)

    Dikstra, B.J.; Farley, J.M.; Scruton, G.

    1990-01-01

    Major developments in ultrasonic techniques, equipment and systems for automated inspection have lead, over a period of about ten years, to the regular application of sophisticated computer-controlled systems during the manufacture of nuclear reactor pressure vessels. Ten years ago the use of procedures defined in a code such as ASME XI might have been considered sufficient, but it is now necessary, as was demonstrated by the results of the UKAEA defect detection trials and the PISC II trials, to apply more comprehensive arrays of probes and higher test sensitivities. The ultrasonic techniques selected are demonstrated to be adequate by modelling or test-block exercises, the automated systems applied are subject to stringent quality assurance testing, and very rigorous inspection procedures are used in conjunction with a high degree of automation to ensure reproducibility of inspection quality. The state-of-the-art in automated ultrasonic testing of pressure vessels by Babcock is described. Current developments by the company, including automated flaw recognition, integrated modelling of inspection capability, and the use of electronically scanned variable-angle probes are reviewed. Examples quoted include the automated ultrasonic inspections of the Sizewell B pressurized water reactor vessel. (author)

  6. Fracture analyses and test of regions with nozzle and hole and curvature influence in nuclear vessel

    International Nuclear Information System (INIS)

    Wang Baisong; Xu Dinggen; Ye Weijuan; Hu Yinbiao; Liang Xingyun; Gu Shaode; Zhou Peiying

    1993-08-01

    For the calculations of stress intensity factor K 1 of surface crack in the regions with nozzle and hole and the curvature influence on nuclear vessel, a improved 3-D collapsed isoparametric singular element with quarter-points was presented. The square root singularity in the vertical planes of crack was derived. The methods of transitional element and calculating K 1 from displacements were extensively used in 3- D case. The SIF K 1 of the corner crack in inner wall of the nozzle of RPV (reactor pressure vessel) for a typical 300 MW nuclear plant was calculated, and it was verified by 3-D photo-elastic test and diffusion of light test. The engineering fracture analysis and evaluation of the outside surface crack in the circular are transitional region of the head flange of RPV are also completed

  7. Investigation of irradiation embrittlement and annealing behaviour of JRQ pressure vessel steel by instrumented impact tests

    Energy Technology Data Exchange (ETDEWEB)

    Valo, M; Rintamaa, R; Nevalainen, M; Wallin, K; Torronen, K [Technical Research Centre of Finland, Espoo (Finland); Tipping, P [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1994-12-31

    Seven series of A533-B type pressure vessel steel specimens irradiated as well as irradiated - annealed - re-irradiated to different fast neutron fluences (up to 5.10{sup 19}/cm{sup 2}) have been tested with a new type of instrumented impact test machine. The radiation embrittlement and the effect of the intermediate annealing was assessed by using the ductile and cleavage fracture initiation toughness. Although the ductile fracture initiation toughness exhibited scatter, the transition temperature shift corresponding to the dynamic cleavage fracture initiation agreed well with the 41 J Charpy-V shift. The results indicate that annealing is beneficial in restoring mechanical properties in an irradiated nuclear pressure vessel steel. (authors). 8 refs., 11 figs., 1 tab.

  8. Application of advanced irradiation analysis methods to light water reactor pressure vessel test and surveillance programs

    International Nuclear Information System (INIS)

    Odette, R.; Dudey, N.; McElroy, W.; Wullaert, R.; Fabry, A.

    1977-01-01

    Inaccurate characterization and inappropriate application of neutron irradiation exposure variables contribute a substantial amount of uncertainty to embrittlement analysis of light water reactor pressure vessels. Damage analysis involves characterization of the irradiation environment (dosimetry), correlation of test and surveillance metallurgical and dosimetry data, and projection of such data to service conditions. Errors in available test and surveillance dosimetry data are estimated to contribute a factor of approximately 2 to the data scatter. Non-physical (empirical) correlation procedures and the need to extrapolate to the vessel may add further error. Substantial reductions in these uncertainties in future programs can be obtained from a more complete application of available damage analysis tools which have been developed for the fast reactor program. An approach to reducing embrittlement analysis errors is described, and specific examples of potential applications are given. The approach is based on damage analysis techniques validated and calibrated in benchmark environments

  9. Thermomechanical Model and Bursting Tests to Evaluate the Risk of Swelling and Bursting of Modified 9Cr-1Mo Steel Steam Generator Tubes during a Sodium-Water Reaction Accident

    Directory of Open Access Journals (Sweden)

    C. Bertrand

    2014-01-01

    Full Text Available The MECTUB code was developed to evaluate the risk of swelling and bursting of Steam Generator (SG tubes. This code deals with the physic of intermediate steam-water leaks into sodium which induce a Sodium-Water Reaction (SWR. It is based on a one-dimensional calculation to describe the thermomechanical behavior of tubes under a high internal pressure and a fast external overheating. The mechanical model of MECTUB is strongly correlated with the kind of the material of the SG tubes. It has been developed and validated by using experiments performed on the alloy 800. A change to tubes made of Modified 9Cr-1Mo steel requires more knowledge of Modified 9Cr-1Mo steel behavior which influences the bursting time at high temperatures (up to 1200°C. Studies have been initiated to adapt the mechanical model and to qualify it for this material. The first part of this paper focuses on the mechanical law modelling (elasticity, plasticity, and creep for Modified 9Cr-1Mo steel and on overheating thermal data. In a second part, the results of bursting tests performed on Modified 9Cr-1Mo tubes in the SQUAT facility of CEA are used to validate the mechanical model of MECTUB for the Modified 9Cr-1Mo material.

  10. Considerations for acoustic emission monitoring of spherical Kevlar/epoxy composite pressure vessels

    Science.gov (United States)

    Hamstad, M. A.; Patterson, R. G.

    1977-01-01

    We are continuing to research the applications of acoustic emission testing for predicting burst pressure of filament-wound Kevlar 49/epoxy pressure vessels. This study has focused on three specific areas. The first area involves development of an experimental technique and the proper instrumentation to measure the energy given off by the acoustic emission transducer per acoustic emission burst. The second area concerns the design of a test fixture in which to mount the composite vessel so that the acoustic emission transducers are held against the outer surface of the composite. Included in this study area is the calibration of the entire test setup including couplant, transducer, electronics, and the instrument measuring the energy per burst. In the third and final area of this study, we consider the number, location, and sensitivity of the acoustic emission transducers used for proof testing composite pressure vessels.

  11. Comparison cosmic ray irradiation simulation and particle beam test on UFFO Burst Alert & Trigger telescope(UBAT) detectors

    DEFF Research Database (Denmark)

    Jeong, H. M.; Jeong, S.; Kim, M. B.

    2017-01-01

    Ultra-Fast Flash Observatory pathfinder(UFFO-p) was launched onboard Lomonosov on 28th of April, 2016, and now is under various types of calibration for detection of Gamma Ray Bursts (GRBs). Since last September UFFO-p has taken X-ray data in space with UFFO Burst Alert &Trigger telescope (UBAT),...

  12. The 4th surveillance testing for Kori unit 3 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-10-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

  13. The 5th surveillance testing for Kori unit 2 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

  14. Neutron stars as X-ray burst sources. II. Burst energy histograms and why they burst

    International Nuclear Information System (INIS)

    Baan, W.A.

    1979-01-01

    In this work we explore some of the implications of a model for X-ray burst sources where bursts are caused by Kruskal-Schwarzschild instabilities at the magnetopause of an accreting and rotating neutron star. A number of simplifying assumptions are made in order to test the model using observed burst-energy histograms for the rapid burster MXB 1730--335. The predicted histograms have a correct general shape, but it appears that other effects are important as well, and that mode competition, for instance, may suppress the histograms at high burst energies. An explanation is ventured for the enhancement in the histogram at the highest burst energies, which produces the bimodal shape in high accretion rate histograms. Quantitative criteria are given for deciding when accreting neutron stars are steady sources or burst sources, and these criteria are tested using the X-ray pulsars

  15. Helium leak testing of a radioactive contaminated vessel under high pressure in a contaminated environment

    International Nuclear Information System (INIS)

    Winter, M.E.

    1996-01-01

    At ANL-W, with the shutdown of EBR-II, R ampersand D has evolved from advanced reactor design to the safe handling, processing, packaging, and transporting spent nuclear fuel and nuclear waste. New methods of processing spent fuel rods and transforming contaminated material into acceptable waste forms are now in development. Storage of nuclear waste is a high interest item. ANL-W is participating in research of safe storage of nuclear waste, with the WIPP (Waste Isolation Pilot Plant) site in New Mexico the repository. The vessel under test simulates gas generated by contaminated materials stored underground at the WIPP site. The test vessel is 90% filled with a mixture of contaminated material and salt brine (from WIPP site) and pressurized with N2-1% He at 2500 psia. Test acceptance criteria is leakage -7 cc/seconds at 2500 psia. The bell jar method is used to determine leakage rate using a mass spectrometer leak detector (MSLD). The efficient MSLD and an Al bell jar replaced a costly, time consuming pressure decay test setup. Misinterpretation of test criterion data caused lengthy delays, resulting in the development of a unique procedure. Reevaluation of the initial intent of the test criteria resulted in leak tolerances being corrected and test efficiency improved

  16. Ex-vessel debris coolability test during severe accident (COTELS project)

    International Nuclear Information System (INIS)

    Ogasawara, H.

    1998-01-01

    The objectives of the COTELS project are for severe accident management, to investigate phenomena of ex-vessel fuel-coolant interactions after reactor pressure vessel (RPV) failure and to investigate molten core-concrete interaction when coolant is injected onto molten debris. The project has being cooperated with the National Nuclear Center in the Republic of Kazakstan from 1994 to 1997 under the sponsorship of the Ministry of International Trade and Industry of Japan. Total programs are composed with the following tests. (1) Test 01 was meant to observe flow mode of falling debris. (2) Test A was meant to investigate phenomena of fuel-coolant interactions when molten debris falls into a coolant pool. (3) Test B/C investigated fuel coolant interactions and molten core-concrete interaction when coolant is injected onto debris. Detail data evaluation is underway. The following results were thus for obtained: (1) It was confirmed in Test 01 series that about 60 kg of UO 2 mixture was completely melted and fallen as a continuous jet. (2) No energetic fuel-coolant interaction was observed both in Test A and B series. (3) Debris in which decay heat was simulated was cooled by water injection in Test C series

  17. Analysis of Accelerometer Data from a Woven Inflatable Creep Burst Test

    Science.gov (United States)

    James, George H.; Grygier, Michael; Selig, Molly M.

    2015-01-01

    Accelerometers were used to montor an inflatable test article during a creep test to failure. The test article experienced impulse events that were classified based on the response of the sensors and their time-dependent manifestation. These impulse events required specialized techniques to process the structural dynamics data. However, certain phenomena were defined as worthy of additional study. An assessment of one phenomena (a frequency near 1000Hz) showed a time dependent frequency and an amplitude that increased significantly near the end of the test. Hence, these observations are expected to drive future understanding of and utility in inflatable space structures.

  18. Remote maintenance of in-vessel components in Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Loesser, G.D.; Heitzenroeder, P.; Kungl, D.; Dylla, H.F.; Cerdan, G.

    1990-01-01

    The Tokamak Fusion Test Reactor (TFTR) will generate a total of 3 x 10 21 neutrons during its planned D-T operational period. A maintenance manipulator has been designed and tested to minimize personnel radiation during in-vessel maintenance activities. Its functions include visual inspection, first-wall tile replacement, cleaning, diagnostics calibrations and leak detection. To meet these objectives, the TFTR maintenance manipulator is required to be operable in the TFTR high vacuum environment, typically -8 torr, ( -6 Pa). Geometrically, the manipulator must extend 180 0 in either direction around the torus to assure complete coverage of the vessel first-wall. The manipulator consists of a movable carriage, and movable articulated link sections which are driven by electrical actuators. The boom has vertical load capacity of 455 kg and lateral load capacity of 46 kg. The boom can either be fitted with a general inspection arm or dextrous slave arms. The general inspection arm is designed to hold the leak detector and an inspection camera; it is capable of rotation along two axes and has a linkage system which permits motion normal to the vacuum vessel wall. All systems except the dextrous slave arms are operable in a vacuum. (author)

  19. Tensile and burst tests in support of the cadmium safety rod failure evaluation

    International Nuclear Information System (INIS)

    Thomas, J.K.

    1992-02-01

    The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of hypothetical LOCA event. Accordingly, an experimental safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. This report discusses confirmatory separate effects tests conducted to support the evaluation of failures observed in the safety rod thermal tests. As part of the failure evaluation, the potential for liquid metal embrittlement (LME) of the safety rod cladding by cadmium (Cd) -- aluminum (Al) solutions was examined. Based on the test conditions, literature data, and U-Bend tests, its was concluded that the SS304 safety rod cladding would not be subject to LME by liquid Cd-Al solutions under conditions relevant to the safety rod thermal tests or gamma heating accident. To confirm this conclusion, tensile tests on SS304 specimens were performed in both air and liquid Cd-Al solutions with the range of strain rates, temperatures, and loading conditions spanning the range relevant to the safety rod thermal tests and gamma heating accident

  20. Testing of plain and fibrous concrete single cavity prestressed concrete reactor vessel models

    International Nuclear Information System (INIS)

    Oland, C.B.

    1985-01-01

    Two single-cavity prestressed concrete reactor vessel (PCRV) models were fabricated and tested to failure to demonstrate the structural response and ultimate pressure capacity of models cast from high-strength concretes. Concretes with design compressive strengths in excess of 70 MPa (10,000 psi) were developed for this investigation. One model was cast from plain concrete and failed in shear at the head region. The second model was cast from fiber reinforced concrete and failed by rupturing the circumferential prestressing at the sidewall of the structure. The tests also demonstrated the capabilities of the liner system to maintain a leak-tight pressure boundary. 3 refs., 4 figs

  1. Three-Dimensional Digital Image Correlation of a Composite Overwrapped Pressure Vessel During Hydrostatic Pressure Tests

    Science.gov (United States)

    Revilock, Duane M., Jr.; Thesken, John C.; Schmidt, Timothy E.

    2007-01-01

    Ambient temperature hydrostatic pressurization tests were conducted on a composite overwrapped pressure vessel (COPV) to understand the fiber stresses in COPV components. Two three-dimensional digital image correlation systems with high speed cameras were used in the evaluation to provide full field displacement and strain data for each pressurization test. A few of the key findings will be discussed including how the principal strains provided better insight into system behavior than traditional gauges, a high localized strain that was measured where gages were not present and the challenges of measuring curved surfaces with the use of a 1.25 in. thick layered polycarbonate panel that protected the cameras.

  2. Preliminary reactor physics calculations for Exxon LWR fuel testing in the power burst facility

    International Nuclear Information System (INIS)

    Olson, W.O.; Nigg, D.W.

    1981-05-01

    The PFB reactor is being considered as an irradiation facility to test LWR fuel rods for Exxon Nuclear Company. Requested test conditions are 18 kW/ft axial peak steady state power in 2.5% initial enrichment, 20,000 MWd/Tu exposed rods. Multigroup transport theory calculations (S/sub n/ and Monte Carlo) showed that this was unattainable in the standard PBF test loop. Thus, a flux multiplier was developed in the form of a Zr-2-clad 0.15-inch thick cylindrical shell of 35% enriched, 88% T.D. UO 2 replacing the flow divider, surrounding the rod within the in-pile tube in PFB. With this flux multiplier installed and assuming an average water density of 0.86 g/cm 3 within the test loop, a Figure of Merit (FOM) for a single-rod test assembly of 0.86 kW/ft-MW +- 5% (at 95% confidence level) was calculated. This FOM is the axial peak linear test rod power per megawatt of reactor power. A reactor power of about 21 megawatts will therefore be required to supply the requested linear test rod axial peak heating rate of 18 kW/ft

  3. Plan on test to failure of a prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Takumi, K.; Nonaka, A.; Umeki, K.; Nagata, K.; Soejima, M.; Yamaura, Y.; Costello, J.F.; Riesemann, W.A. von.; Parks, M.B.; Horschel, D.S.

    1992-01-01

    A summary of the plans to test a prestressed concrete containment vessel (PCCV) model to failure is provided in this paper. The test will be conducted as a part of a joint research program between the Nuclear Power Engineering Corporation (NUPEC), the United States Nuclear Regulatory Commission (NRC), and Sandia National Laboratories (SNL). The containment model will be a scaled representation of a PCCV for a pressurized water reactor (PWR). During the test, the model will be slowly pressurized internally until failure of the containment pressure boundary occurs. The objectives of the test are to measure the failure pressure, to observe the mode of failure, and to record the containment structural response up to failure. Pre- and posttest analyses will be conducted to forecast and evaluate the test results. Based on these results, a validated method for evaluating the structural behavior of an actual PWR PCCV will be developed. The concepts to design the PCCV model are also described in the paper

  4. Fabrication and mechanical test data for the four 6-inch-thick intermediate test vessels made from steel plate for the Heavy Section Steel Program

    International Nuclear Information System (INIS)

    Childress, C.E.

    1976-01-01

    The HSST Program has among its goals the objective of demonstrating the capability to predict safe behavior of thick-walled pressure vessels containing flaws of known dimensions under frangible, transitional, and tough loading regimes. To accomplish these objectives the program is conducting a series of tests involving 6-in.-thick pressure vessels which will serve as test specimens for assisting in the characterization of failure under these loading conditions. Among the vessels a number of parameters, such as weld type, weld location, flaw size and shape, and test temperature and pressure, will be selectively varied to show that a rationale exists for dealing with the varied stress and metallurgical states which normally exist in commercial nuclear reactor vessels. Each vessel will serve as a go, no-go determination of critical flaw size for a specific set of test parameters. Item 4 of the previous issues in this series covers the fabrication details of the first six 6-in.-thick test vessels, which were fabricated from ASTM A-508 Cl 2 forging materials. This report covers the fabrication details of four additional 6-in.-thick intermediate test vessels having shell courses fabricated from ASTM A-533 Gr B Cl 1 plate. The remaining components were made from forgings. Essentially this report is a continuation of ORNL-TM-4351; it describes the manufacturing details of the individual parts and their ultimate assembly into finished vessels. Details concerning chemical composition and mechanical and nondestructive test data are presented

  5. Introduction to the modified TROI test facility for fuel coolant interaction under a submerged reactor vessel

    International Nuclear Information System (INIS)

    Na, Young Su; Hong, Seong-Wan; Song, Jin Ho; Hong, Seong-Ho

    2014-01-01

    The molten Fuel-Coolant Interaction (FCI) can threaten the integrity of the reactor cavity under a severe accident. A steam explosion can be occurred by the rapid energy transfer in the high-temperature corium melt jet penetrating into water, which makes the dynamic load applying to the surrounding structure. Before a steam explosion, the corium melt jet breaks into small-sized particles, and the steam is generated continuously by the film boiling on the hot surface of the melt contacting with water. The premixing phase consisting of the corium melt, water, and steam can determine the intensity of the steam explosion. Unfortunately, the previous experimental studies on the FCI phenomena have carried out under a free fall of the corium melt jet in a gas phase before interacting with water. The previous TROI (Test for Real cOrium Interaction with water) test facility, that is a well-known test facility for the FCI phenomena in the world, has observed a steam explosion under a free fall of a corium melt jet in a gas phase before contacting a coolant since 2000, which is changing to simulate the FCI phenomena under a submerged reactor vessel. This study introduces the modified TROI test facility as shown in Fig. 1 and the considerations for the experiment with success. The previous TROI test facility, that has observed the molten Fuel-Coolant Interaction (FCI) with a free fall of the prototypic corium melt in a gas phase before contacting a coolant, was modified to simulate the FCI phenomena under a submerged reactor vessel for the assessment of the In-Vessel Retention (IVR) concept, i.e., without a free-fall distance of the corium melt before contacting water. The superheated prototypic corium melt created by the cold crucible melting method moves on a releasing valve newly installed just above the water level in the interaction vessel. The corium melt will stay on a releasing valve in less than 0.2 seconds to reduce heat loss for preventing the solidification, and

  6. 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

  7. Round Robin Posttest analysis of a 1/10-scale Steel Containment Vessel Model Test

    International Nuclear Information System (INIS)

    Komine, Kuniaki; Konno, Mutsuo

    1999-01-01

    NUPEC and U.S. Nuclear Regulatory Commission (USNRC) have been jointly sponsoring 'Structural Behavior Test' at Sandia National Laboratory (SNL) in Cooperative Containment Research Program'. As one of the test, a test of a mixed scaled SCV model with 1/10 in the geometry and 1/4 in the shell thickness. Round Robin analyses of a 1/10-scale Steel Containment Vessel (SCV) Model Test were carried out to obtain an adequate analytical method among seven organizations belonged to five countries in the world. As one of sponsor, Nuclear Power Engineering Corporation (NUPEC) filled the important role of a posttest analysis of SCV model. This paper describes NUPEC's analytical results in the round robin posttest analysis. (author)

  8. Round Robin Posttest analysis of a 1/10-scale Steel Containment Vessel Model Test

    Energy Technology Data Exchange (ETDEWEB)

    Komine, Kuniaki [Nuclear Power Engineering Corp., Tokyo (Japan); Konno, Mutsuo

    1999-07-01

    NUPEC and U.S. Nuclear Regulatory Commission (USNRC) have been jointly sponsoring 'Structural Behavior Test' at Sandia National Laboratory (SNL) in Cooperative Containment Research Program'. As one of the test, a test of a mixed scaled SCV model with 1/10 in the geometry and 1/4 in the shell thickness. Round Robin analyses of a 1/10-scale Steel Containment Vessel (SCV) Model Test were carried out to obtain an adequate analytical method among seven organizations belonged to five countries in the world. As one of sponsor, Nuclear Power Engineering Corporation (NUPEC) filled the important role of a posttest analysis of SCV model. This paper describes NUPEC's analytical results in the round robin posttest analysis. (author)

  9. Surveillance tests for light-water cooled nuclear power reactor vessels in IMEF

    International Nuclear Information System (INIS)

    Choo, Yong-Sun; Ahn, Sang-Bok; Park, Dae-Gyu; Jung, Yang-Hong; Yoo, Byung-Ok; Oh, Wan-Ho; Baik, Seung-Je; Koo, Dae-Seo; Lee, Key-Soon

    1999-01-01

    The surveillance tests for light-water cooled nuclear power reactor vessels were established to monitor the radiation-induced changes in the mechanical properties of ferritic materials in the beltline according to US NRC 10 CFR 50 App. G, US NRC RG1.99-rev.2, ASTM E185-82 and E185-94 in Irradiated Materials Examination Facility(IMEF). The surveillance capsule was transported from NPPs pool sites to KAERI IMEF by using a shipping cask. The capsule was cut and dismantled by capsule cutting machine and milling machine in M2 hot cell. Charpy tests and tension tests were performed in M5a and M5b hot cells respectively. Especially the EPMA located at hot lab was used to analyze the Ni and Cu wt% composition of base metal and weld for predicting the adjusted reference temperature(ART). The established process and test results were summarized in this paper. (author)

  10. Test results on direct containment heating by high-pressure melt ejection into the Surtsey vessel: The TDS test series

    International Nuclear Information System (INIS)

    Allen, M.D.; Blanchat, T.K.; Pilch, M.M.

    1994-08-01

    The Technology Development and Scoping (TDS) test series was conducted to test and develop instrumentation and procedures for performing steam-driven, high-pressure melt ejection (HPME) experiments at the Surtsey Test Facility to investigate direct containment heating (DCH). Seven experiments, designated TDS-1 through TDS-7, were performed in this test series. These experiments were conducted using similar initial conditions; the primary variable was the initial pressure in the Surtsey vessel. All experiments in this test series were performed with a steam driving gas pressure of ≅ 4 MPa, 80 kg of lumina/iron/chromium thermite melt simulant, an initial hole diameter of 4.8 cm (which ablated to a final hole diameter of ≅ 6 cm), and a 1/10th linear scale model of the Surry reactor cavity. The Surtsey vessel was purged with argon ( 2 ) to limit the recombination of hydrogen and oxygen, and gas grab samples were taken to measure the amount of hydrogen produced

  11. Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E706 (IIIC)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This test method describes the concept and use of helium accumulation for neutron fluence dosimetry for reactor vessel surveillance. Although this test method is directed toward applications in vessel surveillance, the concepts and techniques are equally applicable to the general field of neutron dosimetry. The various applications of this test method for reactor vessel surveillance are as follows: 1.1.1 Helium accumulation fluence monitor (HAFM) capsules, 1.1.2 Unencapsulated, or cadmium or gadolinium covered, radiometric monitors (RM) and HAFM wires for helium analysis, 1.1.3 Charpy test block samples for helium accumulation, and 1.1.4 Reactor vessel (RV) wall samples for helium accumulation. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  12. Preliminary calculation with code CONTEMPT-LT for spray cooling tests with JAERI model containment vessel

    International Nuclear Information System (INIS)

    Tanaka, Mitsugu

    1978-01-01

    LWR plants have a containment spray system to reduce the escape of radioactive material to the environment in a loss-of-coolant accident (LOCA) by washing out fission products, especially radioiodine, and condensing the steam to lower the pressure. For carrying out the containment spray tests, pressure and temperature behaviour of the JAERI Model Containment Vessel in spray cooling has been calculated with computer program CONTEMPT-LT. The following could be studied quantitatively: (1) pressure and temperature raise rates for steam addition rate and (2) pressure fall rate for spray flow rate and spray heat transfer efficiency. (auth.)

  13. Stable crack growth during over stressing or proof testing of pressure vessels

    International Nuclear Information System (INIS)

    Njo, D.H.

    1985-07-01

    In an earlier study several years ago an attempt has been made by a study group in Switzerland formed by representatives of utilities, manufacturers, inspection agency and the licensing authority to evaluate systematically the advantages and disadvantages of a proof test for a reactor pressure vessel (RPV) taking into account the different aspects involved. In this study, the present day practice in the requirements regarding proof testing in several countries with nuclear power plants were assessed. Using the method of fracture mechanics, taking into account the proof test conditions and material behaviour at hand, qualified quantitative statements were then sought for. One of the important findings of the study was that a sufficiently accurate determination of stable crack growth (SCG) during proof testing is a necessary precondition for a qualified quantitative statement. At that time this requirement could not be met. The aim of the present exercise of collecting available information on SCG was twofold: to get an overview of the present day requirements on pre-service and in-service proof testing of reactor pressure vessels in different countries with NPPs; to collect data on stable crack growth due to over stressing during proof tests to be used in the quantitative assessment, using fracture mechanics methods, and the potential benefit, if any, of proof tests. In the framework of the present study, the originally stipulated aims could not be achieved. It turned out to be a very complex and difficult problem, where various important aspects are little known and could not be assessed without further comprehensive theoretical and experimental investigations

  14. Dormancy release of Norway spruce under climatic warming: testing ecophysiological models of bud burst with a whole-tree chamber experiment.

    Science.gov (United States)

    Hänninen, Heikki; Slaney, Michelle; Linder, Sune

    2007-02-01

    Ecophysiological models predicting timing of bud burst were tested with data gathered from 40-year-old Norway spruce (Picea abies (L.) Karst.) trees growing in northern Sweden in whole-tree chambers under climatic conditions predicted to prevail in 2100. Norway spruce trees, with heights between 5 and 7 m, were enclosed in individual chambers that provided a factorial combination of ambient (365 micromol mol-1) or elevated (700 micromol mol-1) atmospheric CO2 concentration, [CO2], and ambient or elevated air temperature. Temperature elevation above ambient ranged from +2.8 degrees C in summer to +5.6 degrees C in winter. Compared with control trees, elevated air temperature hastened bud burst by 2 to 3 weeks, whereas elevated [CO2] had no effect on the timing of bud burst. A simple model based on the assumption that bud rest completion takes place on a fixed calendar day predicted timing of bud burst more accurately than two more complicated models in which bud rest completion is caused by accumulated chilling. Together with some recent studies, the results suggest that, in adult trees, some additional environmental cues besides chilling are required for bud rest completion. Although it appears that these additional factors will protect trees under predicted climatic warming conditions, increased risk of frost damage associated with earlier bud burst cannot be ruled out. Inconsistent and partially anomalous results obtained in the model fitting show that, in addition to phenological data gathered under field conditions, more specific data from growth chamber and greenhouse experiments are needed for further development and testing of the models.

  15. Manufacturing and material properties of forgings for reactor pressure vessel of high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Sato, I.; Suzuki, K.

    1994-01-01

    For the reactor pressure vessel (RPV) of high temperature engineering test reactor (HTTR) which has been developed by Japan Atomic Energy Research Institute (JAERI), 2 1/4Cr-1Mo steel is used first in the world. Material confirmation test has been carried out to demonstrate good applicability of forged low Si 2 1/4Cr-1Mo steel to the RPV of HTTR. Recently, JSW has succeeded in the manufacturing of large size ring forgings and large size forged cover dome integrated with nozzles for stand pipe for the RPV. This paper describes the results of the material confirmation test as well as the manufacturing and material properties of the large forged cover dome integrated with nozzles for stand pipe. (orig.)

  16. Final Report for the Testing of the Y-12 Criticality Accident Alarm System Detectors at the Godiva IV Burst Reactor (IER-443)

    Energy Technology Data Exchange (ETDEWEB)

    Scorby, John C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hickman, David [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hudson, Becka [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Beller, Tim [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Goda, Joetta [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Haught, Chris [Y-12 National Security Complex, Oak Ridge, TN (United States); Woodrow, Christopher [Y-12 National Security Complex, Oak Ridge, TN (United States); Ward, Dann [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilson, Chris [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Clark, Leo [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom)

    2018-01-05

    This report documents the experimental conditions and final results for the performance testing of the Y-12 Criticality Accident Alarm System (CAAS) detectors at the Godiva IV Burst Reactor at the National Criticality Experimental Research Center (NCERC) at the Nevada National Security Site (NNSS). The testing followed a previously issued test plan and was conducted during the week of July 17, 2017, with completion on Thursday July 20. The test subjected CAAS detectors supplied by Y-12 to very intense and short duration mixed neutron and gamma radiation fields to establish compliance to maximum radiation and minimum pulse width requirements. ANSI/ANS- 8.3.1997 states that the “system shall be sufficiently robust as to actuate an alarm signal when exposed to the maximum radiation expected”, which has been defined at Y-12, in Documented Safety Analyses (DSAs), to be a dose rate of 10 Rad/s. ANSI/ANS-8.3.1997 further states that “alarm actuation shall occur as a result of a minimum duration transient” which may be assumed to be 1 msec. The pulse widths and dose rates provided by each burst during the test exceeded those requirements. The CAAS detectors all provided an immediate alarm signal and remained operable after the bursts establishing compliance to the requirements and fitness for re-deployment at Y-12.

  17. Gamma irradiation testing of prototype ITER in-vessel magnetic pick-up coils

    International Nuclear Information System (INIS)

    Vermeeren, Ludo; Leysen, Willem

    2013-01-01

    Highlights: ► We tested five prototype ITER in-vessel coils up to a gamma dose of 72 MGy. ► Before and after irradiation thermal tests were also performed from 30 °C till 130 °C. ► The continuity resistances and the insulation resistances were continuously monitored. ► The observed behavior of all coils was satisfactory in all conditions. ► For the further design the mechanical robustness should be taken into account. -- Abstract: To fulfill the requirements for ITER in-vessel magnetic diagnostics, several coil prototypes have been developed, aiming at minimizing the disturbing effects of temperature gradients and radiation induced phenomena. As a first step in the radiation resistance testing of these prototypes, an in-situ high dose rate gamma radiation test on a selection of prototypes was performed. The aim of this test was to get a first experimental feedback regarding the behavior of the pick-up coil prototypes under radiation. Five prototypes (a coil wound with glass-insulated copper wire, two LTCC coils and two HTCC coils) were irradiated at a dose rate of 46 kGy/h up to a total dose of 72 MGy and at a temperature of 50 °C. During the irradiation, the continuity resistances and the insulation resistances were continuously measured. Before and after irradiation reference data were recorded as a function of temperature (from 30 °C to 130 °C). This paper includes the results of the temperature and irradiation tests and a discussion of the behavior of the prototype coils in terms of electrical and mechanical properties

  18. Detecting pipe bursts by monitoring water demand

    NARCIS (Netherlands)

    Bakker, M.; Vreeburg, J.H.G.; Van der Roer, M.; Sperber, V.

    2012-01-01

    An algorithm which compares measured and predicted water demands to detect pipe bursts was developed and tested on three data sets of water demand and reported pipe bursts of three years. The algorithm proved to be able to detect bursts where the water loss exceeds 30% of the average water demand in

  19. Integrated leak rate test of the FFTF [Fast Flux Test Facility] containment vessel

    International Nuclear Information System (INIS)

    Grygiel, M.L.; Davis, R.H.; Polzin, D.L.; Yule, W.D.

    1987-04-01

    The third integrated leak rate test (ILRT) performed at the Fast Flux Test Facility (FFTF) demonstrated that effective leak rate measurements could be obtained at a pressure of 2 psig. In addition, innovative data reduction methods demonstrated the ability to accurately account for diurnal variations in containment pressure and temperature. Further development of methods used in this test indicate significant savings in the time and effort required to perform an ILRT on Liquid Metal Reactor Systems with consequent reduction in test costs

  20. The Kolmogorov-Smirnov test for three redshift distributions of long gamma-ray bursts in the Swift Era

    International Nuclear Information System (INIS)

    Dong Yunming; Lu Tan

    2009-01-01

    We investigate redshift distributions of three long burst samples, with the first sample containing 131 long bursts with observed redshifts, the second including 220 long bursts with pseudo-redshifts calculated by the variability-luminosity relation, and the third including 1194 long bursts with pseudo-redshifts calculated by the lag-luminosity relation, respectively. In the redshift range 0-1 the Kolmogorov-Smirnov probability of the observed redshift distribution and that of the variability-luminosity relation is large. In the redshift ranges 1-2, 2-3, 3-6.3 and 0-37, the Kolmogorov-Smirnov probabilities of the redshift distribution from lag-luminosity relation and the observed redshift distribution are also large. For the GRBs, which appear both in the two pseudo-redshift burst samples, the KS probability of the pseudo-redshift distribution from the lag-luminosity relation and the observed reshift distribution is 0.447, which is very large. Based on these results, some conclusions are drawn: i) the V-L iso relation might be more believable than the τ-L iso relation in low redshift ranges and the τ-L iso relation might be more real than the V-Liso relation in high redshift ranges; ii) if we do not consider the redshift ranges, the τ-L iso relation might be more physical and intrinsical than the V-L i so relation. (research papers)

  1. Fracture behaviour assessment of a flawed pressure vessel in the hydro-test

    Energy Technology Data Exchange (ETDEWEB)

    Sarkimo, M; Rintamac, R

    1988-12-31

    This document deals with the fracture properties of a flawed pressure vessel. The experiment was carried out within the Nordic Countries on a vessel in a Finnish refinery. The instrumentation used included acoustic emission. Some results are provided. (TEC).

  2. Preliminary fracture analysis on the integrity of HSST intermediate test vessels

    International Nuclear Information System (INIS)

    Zahoor, A.; Paris, P.C.

    1980-01-01

    Whenever pressure in the vessels is such that the vessel has not yielded the indication is for stable crack growth. With higher pressures which cause full thickness yielding there is unstable crack growth

  3. Relocation work of temporary thermocouples for measuring the vessel cooling system in the safety demonstration test

    International Nuclear Information System (INIS)

    Shimazaki, Yosuke; Shinohara, Masanori; Ono, Masato; Yanagi, Shunki; Tochio, Daisuke; Iigaki, Kazuhiko

    2012-05-01

    It is necessary to confirm that the temperature of water cooling panel of the vessel cooling system (VCS) is controlled under the allowable working temperature during the safety demonstration test because the water cooling panel temperature rises due to stop of cooling water circulation pumps. Therefore, several temporary thermocouples are relocated to the water cooling panel near the stabilizers of RPV and the side cooling panel outlet ring header of VCS in order to observe the temperature change of VCS. The relocated thermocouples can measure the temperature change with starting of the cooling water circulation pumps of VCS. So it is confirmed that the relocated thermocouples can observe the VCS temperature change in the safety demonstration test. (author)

  4. Report on material and fabrication tests of the KUHFR core vessel

    International Nuclear Information System (INIS)

    Yoshida, H.; Kozuka, T.; Achiwa, N.; Mitani, S.; Kawano, S.; Araki, Y.; Shibata, T.

    1983-01-01

    For the material of the cylindrical reactor core vessel of the Kyoto University High Flux Reactor (KUHFR), A6061 alloy is selected because the aged state of the alloy is known to show the highest resistance against void swelling due to high-dose irradiation. The fabrication possibility of the large-scale tubes is also tested because the sizes (40 cmdiameter and 43 cmdiameter x 960 cm with a thickness of 10 mm for the inner- and outer-tubes, respectively) are just over the largest limit of the conventional factory fabrication. The results are summarized as follows. (1) From an ingot of A6061 alloy a raw inner-tube is hot-extruded by the 3,000 ton press machine. The shape of the extruded tubes is effectively corrected by stretch forming and other special methods. (2) The real scale tubes are heat-treated under the various conditions (T1, T4 and T6) and their size changes are measured just after the every heat-treatment. (3) The hydropressure for a pipe prepared by welding from an aged-tube shows a fairly uniform strain distribution and the breaking initiation at the reasonable pressure in the welded part. (4) Each of the welded specimens prepared using three kinds of welding rods shows sufficient strength in both of bending and tensile test for the JIS standard. Their microstructures correspond to the result of the mechanical tests for each welded specimen. The confidence for the fabrication possibility of the real core vessel has been given through the present tests. (author)

  5. PBF [Power Burst Facility] severe fuel damage test 1-1: Volume 2, Test results report, Appendices A through I

    International Nuclear Information System (INIS)

    Martinson, Z.R.; Petti, D.A.; Cook, B.A.

    1986-10-01

    This report provides information on: fuel rod characteristics; instrumentation identification, location, and performance; effluent sampling and monitoring system; bundle power; test SFD 1-1 data qualification, uncertainties, and data plots; postirradiation examinations; chemical kinetics predictions; SCDAP analysis model; and coolant level measurements

  6. BETHSY 9.1b Test Calculation with TRACE Using 3D Vessel Component

    International Nuclear Information System (INIS)

    Berar, O.; Prosek, A.

    2012-01-01

    Recently, several advanced multidimensional computational tools for simulating reactor system behaviour during real and hypothetical transient scenarios were developed. One of such advanced, best-estimate reactor systems codes is TRAC/RELAP Advanced Computational Engine (TRACE), developed by the U.S. Nuclear Regulatory Commission. The advanced TRACE comes with a graphical user interface called SNAP (Symbolic Nuclear Analysis Package). It is intended for pre- and post-processing, running codes, RELAP5 to TRACE input deck conversion, input deck database generation etc. The TRACE code is still not fully development and it will have all the capabilities of RELAP5. The purpose of the present study was therefore to assess the 3D capability of the TRACE on BETHSY 9.1b test. The TRACE input deck was semi-converted (using SNAP and manual corrections) from the RELAP5 input deck. The 3D fluid dynamics within reactor vessel was modelled and compared to 1D fluid dynamics. The 3D calculation was compared both to TRACE 1D calculation and RELAP5 calculation. Namely, the geometry used in TRACE is basically the same, what gives very good basis for the comparison of the codes. The only exception is 3D reactor vessel model in case of TRACE 3D calculation. The TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 4 were used for calculations. The BETHSY 9.1b test (International Standard Problem no. 27 or ISP-27) was 5.08 cm equivalent diameter cold leg break without high pressure safety injection and with delayed ultimate procedure. BETHSY facility was a 3-loop replica of a 900 MWe FRAMATOME pressurized water reactor. For better presentation of the calculated physical phenomena and processes, an animation model using SNAP was developed. In general, the TRACE 3D code calculation is in good agreement with the BETHSY 9.1b test. The TRACE 3D calculation results are as good as or better than the RELAP5 calculated results. Also, the TRACE 3D calculation is not significantly different from TRACE 1D

  7. Static and dynamic analyses on the MFTF [Mirror Fusion Test Facility]-B Axicell Vacuum Vessel System: Final report

    International Nuclear Information System (INIS)

    Ng, D.S.

    1986-09-01

    The Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory (LLNL) is a large-scale, tandem-mirror-fusion experiment. MFTF-B comprises many highly interconnected systems, including a magnet array and a vacuum vessel. The vessel, which houses the magnet array, is supported by reinforced concrete piers and steel frames resting on an array of foundations and surrounded by a 7-ft-thick concrete shielding vault. The Pittsburgh-Des Moines (PDM) Corporation, which was awarded the contract to design and construct the vessel, carried out fixed-base static and dynamic analyses of a finite-element model of the axicell vessel and magnet systems, including the simulation of various loading conditions and three postulated earthquake excitations. Meanwhile, LLNL monitored PDM's analyses with modeling studies of its own, and independently evaluated the structural responses of the vessel in order to define design criteria for the interface members and other project equipment. The assumptions underlying the finite-element model and the behavior of the axicell vessel are described in detail in this report, with particular emphasis placed on comparing the LLNL and PDM studies and on analyzing the fixed-base behavior with the soil-structure interaction, which occurs between the vessel and the massive concrete vault wall during a postulated seismic event. The structural members that proved sensitive to the soil effect are also reevaluated

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  10. Burst Pressure Failure of Titanium Tanks Damaged by Secondary Plumes from Hypervelocity Impacts on Aluminum Shields

    Science.gov (United States)

    Nahra, Henry; Ghosn, Louis; Christiansen, Eric; Davis, B. Alan; Keddy, Chris; Rodriquez, Karen; Miller, Joshua; Bohl, William

    2011-01-01

    Metallic pressure tanks used in space missions are inherently vulnerable to hypervelocity impacts from micrometeoroids and orbital debris; thereby knowledge of impact damage and its effect on the tank integrity is crucial to a spacecraft risk assessment. This paper describes tests that have been performed to assess the effects of hypervelocity impact (HVI) damage on Titanium alloy (Ti-6Al-4V) pressure vessels burst pressure and characteristics. The tests consisted of a pair of HVI impact tests on water-filled Ti-6Al-4V tanks (water being used as a surrogate to the actual propellant) and subsequent burst tests as well as a burst test on an undamaged control tank. The tanks were placed behind Aluminum (Al) shields and then each was impacted with a 7 km/s projectile. The resulting impact debris plumes partially penetrated the Ti-6Al-4V tank surfaces resulting in a distribution of craters. During the burst tests, the tank that failed at a lower burst pressure did appear to have the failure initiating at a crater site with observed spall cracks. A fracture mechanics analysis showed that the tanks failure at the impact location may have been due to a spall crack that formed upon impact of a fragmentation on the Titanium surface. This result was corroborated with a finite element analysis from calculated Von-Mises and hoop stresses.

  11. Critical Heat Flux Experiments on the Reactor Vessel Wall Using 2-D Slice Test Section

    International Nuclear Information System (INIS)

    Jeong, Yong Hoon; Chang, Soon Heung; Baek, Won-Pil

    2005-01-01

    The critical heat flux (CHF) on the reactor vessel outer wall was measured using the two-dimensional slice test section. The radius and the channel area of the test section were 2.5 m and 10 cm x 15 cm, respectively. The flow channel area and the heater width were smaller than those of the ULPU experiments, but the radius was greater than that of the ULPU. The CHF data under the inlet subcooling of 2 to 25 deg. C and the mass flux 0 to 300 kg/m 2 .s had been acquired. The measured CHF value was generally slightly lower than that of the ULPU. The difference possibly comes from the difference of the test section material and the thickness. However, the general trend of CHF according to the mass flux was similar with that of the ULPU. The experimental CHF data were compared with the predicted values by SULTAN correlation. The SULTAN correlation predicted well this study's data only for the mass flux higher than 200 kg/m 2 .s, and for the exit quality lower than 0.05. The local condition-based correlation was developed, and it showed good prediction capability for broad quality (-0.01 to 0.5) and mass flux ( 2 .s) conditions with a root-mean-square error of 2.4%. There were increases in the CHF with trisodium phosphate-added water

  12. Ultrasonic in-service testing of pressure vessel bodies of nuclear power reactors

    International Nuclear Information System (INIS)

    Obraz, J.

    1978-01-01

    In-service ultrasonic testing of reactor pressure vessels is described using a system of probes for simultaneous testing of material or weld joint thicknesses. The signal is transmitted from a common output via a 30 m long cable to electronic evaluation equipment. The methods are described of ultrasonic detection of fatigue cracks. The static calculation of the dependence of echo amplitudes on crack orientation and the dynamic calculation of the crack orientation effect are described for the indirect reflection technique. In testing, angular probes with gap-type acoustic coupling operating at a frequency of 2 MHz were preferably used. For detecting planar defects of more than 10 mm in size inclined by more than +-10deg probes operating at a frequency of 1 MHz were more advantageous. The direct reflection technique is suitable for detecting defects near the surface (10 to 20 mm) and for cases when the indirect reflection technique cannot be used. For this technique a focusing probe operating at a frequency of 2 MHz is suitable. The strong dependence of the echo amplitude on the crack depth is a disadvantage of the technique. Defects near the surface, i.e., immediately under cladding are best detected by means of a double probe transmitting transversal waves at an angle of 60deg. Experimental measurements were carried out on materials with artificial defects of the type of bores with flat bottom. (J.P.)

  13. Laparoscopic prototype for optical sealing of renal blood vessels

    Science.gov (United States)

    Hardy, Luke A.; Hutchens, Thomas C.; Larson, Eric R.; Gonzalez, David A.; Chang, Chun-Hung; Nau, William H.; Fried, Nathaniel M.

    2017-02-01

    Energy-based, radiofrequency and ultrasonic devices provide rapid sealing of blood vessels during laparoscopic procedures. We are exploring infrared lasers as an alternative for vessel sealing with less collateral thermal damage. Previous studies demonstrated vessel sealing in an in vivo porcine model using a 1470-nm laser. However, the initial prototype was designed for open surgery and featured tissue clasping and light delivery mechanisms incompatible with laparoscopic surgery. In this study, a laparoscopic prototype similar to devices in surgical use was developed, and tests were conducted on porcine renal blood vessels. The 5-mm-OD prototype featured a traditional Maryland jaw configuration. Laser energy was delivered through a 550-μm-core fiber and side-delivery from the lower jaw, with beam dimensions of 18-mm-length x 1.2-mm-width. The 1470-nm diode laser delivered 68 W with 3 s activation time. A total of 69 porcine renal vessels with mean diameter of 3.3 +/- 1.7 mm were tested, ex vivo. Vessels smaller than 5 mm were consistently sealed (48/51) with burst pressures greater than malignant hypertension blood pressure (180 mmHg), averaging 1038 +/- 474 mmHg. Vessels larger than 5 mm were not consistently sealed (6/18), yielding burst pressures of only 174 +/- 221 mmHg. Seal width, thermal damage zone, and thermal spread averaged 1.7 +/- 0.8, 3.4 +/- 0.7, and 1.0 +/- 0.4 mm. A novel optical laparoscopic prototype with 5-mm- OD shaft integrated within a standard Maryland jaw design consistently sealed vessels less than 5 mm with minimal thermal spread. Further in vivo studies are planned to test performance across a variety of vessels and tissues.

  14. Nuclear reactor pressure vessel integrity insurance by crack arrestability evaluation using load from CVN tests

    International Nuclear Information System (INIS)

    Fabry, A.

    1997-01-01

    The present work is undertaken in the framework of nuclear reactor pressure vessel (RPV) surveillance and aims at revisiting the crack arrest approach to structural integrity insurance. This approach, performed under normal plant operation conditions, can also offer an attractive alternative to the crack initiation philosophy promoted for accidental analysis. To this end, an accidental conservative, cost effective and robust methodology is forwarded and demonstrated: it makes use of the crack arrest information contained in the instrumented Charpy V-notch impact test and/or in the shear fracture appearance of broken samples. Particular attention is paid to the appraisal of uncertainties and the related safety margin. The resulting capability is placed in perspective with the state-of-the-art crack initiation methodology based on the slow bend testing of recracked specimens, presently under standardization world-wide. The investigation leads to highlight three conceptual weaknesses of current enfgineering and regulatory practices. Improved crack arrestability evaluation emerges as an optimal approach to insure safe PWR operation up to design end-of-life and beyond

  15. Nuclear reactor pressure vessel integrity insurance by crack arrestability evaluation using load from CVN tests

    Energy Technology Data Exchange (ETDEWEB)

    Fabry, A.

    1997-10-15

    The present work is undertaken in the framework of nuclear reactor pressure vessel (RPV) surveillance and aims at revisiting the crack arrest approach to structural integrity insurance. This approach, performed under normal plant operation conditions, can also offer an attractive alternative to the crack initiation philosophy promoted for accidental analysis. To this end, an accidental conservative, cost effective and robust methodology is forwarded and demonstrated: it makes use of the crack arrest information contained in the instrumented Charpy V-notch impact test and/or in the shear fracture appearance of broken samples. Particular attention is paid to the appraisal of uncertainties and the related safety margin. The resulting capability is placed in perspective with the state-of-the-art crack initiation methodology based on the slow bend testing of recracked specimens, presently under standardization world-wide. The investigation leads to highlight three conceptual weaknesses of current enfgineering and regulatory practices. Improved crack arrestability evaluation emerges as an optimal approach to insure safe PWR operation up to design end-of-life and beyond.

  16. High-performance fiber/epoxy composite pressure vessels

    Science.gov (United States)

    Chiao, T. T.; Hamstad, M. A.; Jessop, E. S.; Toland, R. H.

    1978-01-01

    Activities described include: (1) determining the applicability of an ultrahigh-strength graphite fiber to composite pressure vessels; (2) defining the fatigue performance of thin-titanium-lined, high-strength graphite/epoxy pressure vessel; (3) selecting epoxy resin systems suitable for filament winding; (4) studying the fatigue life potential of Kevlar 49/epoxy pressure vessels; and (5) developing polymer liners for composite pressure vessels. Kevlar 49/epoxy and graphite fiber/epoxy pressure vessels, 10.2 cm in diameter, some with aluminum liners and some with alternation layers of rubber and polymer were fabricated. To determine liner performance, vessels were subjected to gas permeation tests, fatigue cycling, and burst tests, measuring composite performance, fatigue life, and leak rates. Both the metal and the rubber/polymer liner performed well. Proportionately larger pressure vessels (20.3 and 38 cm in diameter) were made and subjected to the same tests. In these larger vessels, line leakage problems with both liners developed the causes of the leaks were identified and some solutions to such liner problems are recommended.

  17. Germ cells may survive clipping and division of the spermatic vessels in surgery for intra-abdominal testes

    DEFF Research Database (Denmark)

    Thorup, J M; Cortes, Dina; Visfeldt, J

    1999-01-01

    Laparoscopy is a well described modality that provides an accurate visual diagnosis upon which further management of intra-abdominal testes may be based. Laparoscopic ligation of spermatic vessels as stage 1 of the procedure is a natural extension of laparoscopy. A staged approach provides adequate...

  18. Crack propagation on spherical pressure vessels

    International Nuclear Information System (INIS)

    Lebey, J.; Roche, R.

    1975-01-01

    The risk presented by a crack on a pressure vessel built with a ductile steel cannot be well evaluated by simple application of the rules of Linear Elastic Fracture Mechanics, which only apply to brittle materials. Tests were carried out on spherical vessels of three different scales built with the same steel. Cracks of different length were machined through the vessel wall. From the results obtained, crack initiation stress (beginning of stable propagation) and instable propagation stress may be plotted against the lengths of these cracks. For small and medium size, subject to ductile fracture, the resulting curves are identical, and may be used for ductile fracture prediction. Brittle rupture was observed on larger vessels and crack propagation occurred at lower stress level. Preceedings curves are not usable for fracture analysis. Ultimate pressure can be computed with a good accuracy by using equivalent energy toughness, Ksub(1cd), characteristic of the metal plates. Satisfactory measurements have been obtained on thin samples. The risks of brittle fracture may then judged by comparing Ksub(1cd) with the calculated K 1 value, in which corrections for vessel shape are taken into account. It is thus possible to establish the bursting pressure of cracked spherical vessels, with the help of two rules, one for brittle fracture, the other for ductile instability. A practical method is proposed on the basis of the work reported here

  19. Hydraulic Simulation of In-vessel Downstream Effect Test Using MARS-KS Code

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Lee, Joon Soo; Ryu, Seung Hoon [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    In-vessel downstream effect test (IDET) has been required to evaluate the effect of debris on long term core cooling following a loss of coolant accident (LOCA) in support of resolution of Generic Safety Issue (GSI) 191. Head loss induced by debris (fiber and particle) accumulated on prototypical fuel assembly (FA) should be compared with the available driving head to the core for the various combinations of LOCA and Emergency Core Cooling System (ECCS) injection. The actual simulation was conducted using MARS-KS code. Also the influence of small difference in gap size which was found in the actual experiment is evaluated using the present model. A simple model to determine the form loss factors of FA and gap in clean state and the debris laden state is discussed based on basic fluid mechanics. Those form loss factors were applied to the hydraulic simulation of a selected IDET using MARS-KS code. The result indicated that the present model can be applied to IDET simulation. The pressure drop influenced by small difference in gap size can be evaluated by the present model with practical assumption.

  20. Evaluation of burst pressure prediction models for line pipes

    International Nuclear Information System (INIS)

    Zhu, Xian-Kui; Leis, Brian N.

    2012-01-01

    Accurate prediction of burst pressure plays a central role in engineering design and integrity assessment of oil and gas pipelines. Theoretical and empirical solutions for such prediction are evaluated in this paper relative to a burst pressure database comprising more than 100 tests covering a variety of pipeline steel grades and pipe sizes. Solutions considered include three based on plasticity theory for the end-capped, thin-walled, defect-free line pipe subjected to internal pressure in terms of the Tresca, von Mises, and ZL (or Zhu-Leis) criteria, one based on a cylindrical instability stress (CIS) concept, and a large group of analytical and empirical models previously evaluated by Law and Bowie (International Journal of Pressure Vessels and Piping, 84, 2007: 487–492). It is found that these models can be categorized into either a Tresca-family or a von Mises-family of solutions, except for those due to Margetson and Zhu-Leis models. The viability of predictions is measured via statistical analyses in terms of a mean error and its standard deviation. Consistent with an independent parallel evaluation using another large database, the Zhu-Leis solution is found best for predicting burst pressure, including consideration of strain hardening effects, while the Tresca strength solutions including Barlow, Maximum shear stress, Turner, and the ASME boiler code provide reasonably good predictions for the class of line-pipe steels with intermediate strain hardening response. - Highlights: ► This paper evaluates different burst pressure prediction models for line pipes. ► The existing models are categorized into two major groups of Tresca and von Mises solutions. ► Prediction quality of each model is assessed statistically using a large full-scale burst test database. ► The Zhu-Leis solution is identified as the best predictive model.

  1. Evaluation of burst pressure prediction models for line pipes

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Xian-Kui, E-mail: zhux@battelle.org [Battelle Memorial Institute, 505 King Avenue, Columbus, OH 43201 (United States); Leis, Brian N. [Battelle Memorial Institute, 505 King Avenue, Columbus, OH 43201 (United States)

    2012-01-15

    Accurate prediction of burst pressure plays a central role in engineering design and integrity assessment of oil and gas pipelines. Theoretical and empirical solutions for such prediction are evaluated in this paper relative to a burst pressure database comprising more than 100 tests covering a variety of pipeline steel grades and pipe sizes. Solutions considered include three based on plasticity theory for the end-capped, thin-walled, defect-free line pipe subjected to internal pressure in terms of the Tresca, von Mises, and ZL (or Zhu-Leis) criteria, one based on a cylindrical instability stress (CIS) concept, and a large group of analytical and empirical models previously evaluated by Law and Bowie (International Journal of Pressure Vessels and Piping, 84, 2007: 487-492). It is found that these models can be categorized into either a Tresca-family or a von Mises-family of solutions, except for those due to Margetson and Zhu-Leis models. The viability of predictions is measured via statistical analyses in terms of a mean error and its standard deviation. Consistent with an independent parallel evaluation using another large database, the Zhu-Leis solution is found best for predicting burst pressure, including consideration of strain hardening effects, while the Tresca strength solutions including Barlow, Maximum shear stress, Turner, and the ASME boiler code provide reasonably good predictions for the class of line-pipe steels with intermediate strain hardening response. - Highlights: Black-Right-Pointing-Pointer This paper evaluates different burst pressure prediction models for line pipes. Black-Right-Pointing-Pointer The existing models are categorized into two major groups of Tresca and von Mises solutions. Black-Right-Pointing-Pointer Prediction quality of each model is assessed statistically using a large full-scale burst test database. Black-Right-Pointing-Pointer The Zhu-Leis solution is identified as the best predictive model.

  2. Design, fabrication and test of double-wall vacuum vessel for JT-60U

    International Nuclear Information System (INIS)

    Uchikawa, Takashi; Ioki, Kimihiro; Ninomiya, Hiromasa.

    1994-01-01

    A double-wall vacuum vessel was designed and fabricated for JT-60U (an upgraded machine of JT-60), which has a plasma current up to 6 MA and a large plasma volume (100 m 3 ). A new concept of Inconel 625 all-welded structure was adopted to the vessel, that comprises an inner plate, square tubes and an outer plate. The vacuum vessel with a multi-arc D-shaped cross section was fabricated by using hot-sizing press. The electromagnetic and structural analysis has been performed for plasma disruption loads. Dynamic responses of the vessel were measured during plasma disruptions, and the observed displacement had a good agreement with the result of FEM analysis. (author)

  3. Minimum critical crack depths in pressure vessels guidelines for nondestructive testing

    International Nuclear Information System (INIS)

    Crossley, M.R.; Townley, C.H.A.

    1983-09-01

    Estimates of the minimum critical depths which can be expected in high quality vessels designed to certain British and American Code rules are given. A simple means of allowing for fatigue crack growth in service is included. The data which are presented can be used to decide what sensitivity and what reporting levels should be employed during an ultrasonic inspection of a pressure vessel. It is emphasised that the minimum crack depths are those which would be relevant to a vessel in which the material is stressed to its maximum permitted value during operation. Stresses may, in practice, be significantly less than this. Less restrictive inspection standards may be established, if it were considered worthwhile to carry out a detailed stress analysis of the particular vessel under examination. (author)

  4. Hot cell examination on the surveillance capsule of SA 533 cl. 1 reactor pressure vessel (1st test report)

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yong Sun; Jung, Y. H.; Yoo, B. O.; Baik, S. J.; Oh, W. H.; Soong, W. S.; Hong, K. P

    2000-08-01

    The post-irradiated examinations such as impact test, tensile test, composition analysis and etc. were conducted to monitor and to evaluate the radiation-induced changes, so called radiation embrittlement, in the mechanical properties of ferritic materials. Those data should be applied to confirm safety as well as reliability of reactor pressure vessel. The scopes and contents of hot cell examination on the surveillance capsule are as follows; - Capsule transportation, cutting, dismantling and classification - Shim block and Dosimeter cutting and dismantling - Impact test - Tensile test - Composition analysis by EPMA - SEM observation on the fractured surface - Hardness test - Radwaste treatment.

  5. Firefighter's compressed air breathing system pressure vessel development program

    Science.gov (United States)

    Beck, E. J.

    1974-01-01

    The research to design, fabricate, test, and deliver a pressure vessel for the main component in an improved high-performance firefighter's breathing system is reported. The principal physical and performance characteristics of the vessel which were required are: (1) maximum weight of 9.0 lb; (2) maximum operating pressure of 4500 psig (charge pressure of 4000 psig); (3) minimum contained volume of 280 in. 3; (4) proof pressure of 6750 psig; (5) minimum burst pressure of 9000 psig following operational and service life; and (6) a minimum service life of 15 years. The vessel developed to fulfill the requirements described was completely sucessful, i.e., every category of performence was satisfied. The average weight of the vessel was found to be about 8.3 lb, well below the 9.0 lb specification requirement.

  6. Performance test of micro-fission chambers for in-vessel neutron monitoring of ITER

    International Nuclear Information System (INIS)

    Yamauchi, Michinori; Nishitani, Takeo; Ochiai, Kentaro; Morimoto, Yuichi; Hori, Jun-ichi; Ebisawa, Katsuyuki; Kasai, Satoshi

    2002-03-01

    A micro-fission chamber with 12 mg UO 2 and a dummy chamber without uranium were fabricated and the performance was tested. They are designed to be installed inside the vacuum vessel of the compact ITER (ITER-FEAT) for neutron monitoring. The vacuum leak rate of the dummy chamber with MI cable, resistances of chambers between central conductor and outer sheath, and mechanical strength up to 50G acceleration were confirmed to meet the design criteria. The gamma-ray sensitivity was measured for the dummy chamber with the 60 Co gamma-ray irradiation facility at JAERI Takasaki. The output signals for gamma-rays in Campbelling mode were estimated to be less than 0.1% of those by neutrons at the location behind the blanket module in ITER-FEAT. The detector response for 14 MeV neutrons was investigated with the FNS facility. Excellent linearity between count rates, square of Campbelling voltage and neutron fluxes was confirmed in the temperature range from 20degC (room) to 250degC. However, a positive dependence of 14 MeV neutron count rates on temperature was observed, which might be caused by the increase in the pulse height with temperature rise. Effects of a change of surrounding materials were evaluated by the sensitivity measurements of the micro-fission chamber inserted into the shielding blanket mock-up. The sensitivity was enhanced by slow-downed neutrons, which agreed with the calculation result by MCNP-4C code. As a result, it was concluded that the developed micro-fission chamber is applicable for ITER-FEAT. (author)

  7. Germ cells may survive clipping and division of the spermatic vessels in surgery for intra-abdominal testes

    DEFF Research Database (Denmark)

    Thorup, J M; Cortes, D; Visfeldt, J

    1999-01-01

    PURPOSE: Laparoscopy is a well described modality that provides an accurate visual diagnosis upon which further management of intra-abdominal testes may be based. Laparoscopic ligation of spermatic vessels as stage 1 of the procedure is a natural extension of laparoscopy. A staged approach provides...... studied 17 nonpalpable testes in 10 patients 1 year and 7 months to 13(1/2) years old. Results of testicular biopsies of 13 intra-abdominal testes taken at stages 1 and 2 of surgery were available for histological comparison. RESULTS: Median number of spermatogonia per tubular cross section...... of the biopsies taken at stage 2 was slightly lower (0.03) compared to the median number at stage 1 (0.06) of the operation but this difference was not significant (p = 0.2031). CONCLUSIONS: Our study shows that the spermatogonia may survive clipping and division of the spermatic vessels, although the number...

  8. Structural design considerations in the Mirror Fusion Test Facility (MFTF-B) vacuum vessel

    International Nuclear Information System (INIS)

    Vepa, K.; Sterbentz, W.H.

    1981-01-01

    In view of favorable results from the Tandem Mirror Experiment (TMX) also at LLNL, the MFTF project is now being rescoped into a large tandem mirror configuration (MFTF-B), which is the mainline approach to a mirror fusion reactor. This paper concerns itself with the structural aspects of the design of the vessel. The vessel and its intended functions are described. The major structural design issues, especially those influenced by the analysis, are described. The objectives of the finite element analysis and their realization are discussed at length

  9. Ex-vessel corium spreading: results from the VULCANO spreading tests

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Christophe E-mail: christophe.journeau@cea.fr; Boccaccio, Eric E-mail: eric.boccaccio@cea.fr; Brayer, Claude; Cognet, Gerard E-mail: gerard.cognet@cea.fr; Haquet, Jean-Francois E-mail: haquet@eloise.cad.cea.fr; Jegou, Claude E-mail: claude.jegou@cea.fr; Piluso, Pascal E-mail: pascal.piluso@cea.fr; Monerris, Jose E-mail: jose.monerris@cea.fr

    2003-07-01

    In the hypothetical case of a nuclear reactor severe accident, the reactor core could melt and form a mixture, called corium, of highly refractory oxides (UO{sub 2}, ZrO{sub 2}) and metallic or oxidized steel, that could eventually flow out of the vessel and mix with the basemat decomposition products (generally oxides such as SiO{sub 2}, Al{sub 2}O{sub 3}, CaO, Fe{sub 2}O{sub 3}, ...). For some years, the French Atomic Energy Commission (CEA) has launched an R and D program which aimed at providing the tools for improving the mastering of severe accidents. Within this program, the VULCANO experimental facility is operated to perform experiments with prototypic corium (corium of realistic chemical composition including depleted UO{sub 2}). This is coupled with the use of specific high-temperature instrumentation requiring in situ cross calibration. This paper is devoted to the 'spreading experiments' performed in the VULCANO facility, in which the effects of flow and solidification are studied. Due to the complex behavior of corium in the solidification range, an interdisciplinary approach has been used combining thermodynamics of multicomponent mixtures, rheological models of silicic semisolid materials, heat transfer at high temperatures, free-surface flow of a fluid with temperature-dependant properties. Twelve high-temperature spreading tests have been performed and analyzed. The main experimental results are the good spreadability of corium-concrete mixtures having large solidification ranges even with viscous silicic melts, the change of microstructure due to cooling rates, the occurrence of a large thermal contact resistance at the corium-substrate interface, the presence of a steep viscosity gradient at the surface, the transient concrete ablation. Furthermore, the experiments showed the presence of the gaseous inclusions in the melt even without concrete substrate. This gas release is linked to the local oxygen content in the melt which is

  10. Cosmic gamma bursts

    International Nuclear Information System (INIS)

    Ehstulin, I.V.

    1980-01-01

    A brief consideration is being given to the history of cosmic gamma burst discovery and modern knowledge of their properties. The time dependence of gamma bursts is described and their possible sources are discussed

  11. Joint High Speed Vessel (JHSV) Follow on Operational Test and Evaluation (FOT and E) Report

    Science.gov (United States)

    2015-09-21

    problem by fabricating new lines that included surge pendants . These new lines allow some limited movement of the two, skin-to-skin moored vessels... bridge . Figure 9. SSDG Number 2, USNS Spearhead Figure 10. Flame Face Surface Pictures of SSDG Cylinder Head 15 Figure 11

  12. Vessel grounding in entrance channels: case studies and physical model tests

    CSIR Research Space (South Africa)

    Tulsi, K

    2014-05-01

    Full Text Available . It was found that high speed impacts of 8 to 12 knots at 10° to the channel side slopes have the potential to damage the hull and require enormous tug forces to re-float the grounded vessel....

  13. Simulation test of aerosol generation from vessels in the pre-treatment system of fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Fujine, Sachio; Kitamura, Koichiro; Kihara, Takehiro [Japan Atomic Energy Research Institute (JAERI), Ibaraki-ken (Japan)

    1997-08-01

    Aerosol concentration and droplet size are measured in off-gas of vessel under various conditions by changing off-gas flow rate, stirring air flow rate, salts concentration and temperature of nitrate solution. Aerosols are also measured under evaporation and air-lift operation. 4 refs., 6 figs.

  14. Ultrasonic test results for the reactor pressure vessel of the HTTR. Longitudinal welding line of bottom dome

    International Nuclear Information System (INIS)

    Nojiri, Naoki; Ohwada, Hiroyuki; Kato, Yasushi

    2008-06-01

    This paper describes the inspection method, the measured area, etc. of the ultrasonic test of the in-service inspection (ISI) for welding lines of the reactor pressure vessel of the HTTR and the inspection results of the longitudinal welding line of the bottom dome. The pre-service inspection (PSI) results for estimation of occurrence and progression of defects to compare the ISI results is described also. (author)

  15. Vessel Operating Units (Vessels)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains data for vessels that are greater than five net tons and have a current US Coast Guard documentation number. Beginning in1979, the NMFS...

  16. The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, A., E-mail: alan.peacock@ipp.mpg.de [European Commission c/o Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Girlinger, A. [MAN Diesel and Turbo SE D-94469 Deggendorf (Germany); Vorkoeper, A. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 17491 Greifswald (Germany); Boscary, J.; Greuner, H. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Hurd, F. [European Commission c/o Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Mendelevitch, B.; Pirsch, H.; Stadler, R.; Zangl, G. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany)

    2011-10-15

    320 In-vessel water cooled stainless steel panels, poloidal closure plates and pumping gap panels, covering an area of approximately 100 m{sup 2}, are used in Wendelstein7-X to protect the plasma vessel. The panels are manufactured at Deggendorf, Germany by MAN Diesel and Turbo SE. The panels consist of a laser welded sandwich of stainless steel plates together with a labyrinth of cooling channels and have a complicated geometry to fit the plasma vessel of Wendelstein 7-X. The hydraulic and mechanical stability requirements whilst maintaining the tight tolerances for the shape of the components are very demanding. The panels are designed to operate at up to an average heat load of 100 kW/m{sup 2} and a maximum heat load of 200 kW/m{sup 2} with a water velocity of approximately 2 m s{sup -1}. High heat flux testing of an un-cooled panel at a time averaged load of 200 kW/m{sup 2} for 10 s were successfully performed to support the start up phase of Wendelstein 7-X operation. Extensive testing both during manufacture and after delivery to IPP-Garching demonstrates the suitability of the delivered panels for their purpose.

  17. The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator

    International Nuclear Information System (INIS)

    Peacock, A.; Girlinger, A.; Vorkoeper, A.; Boscary, J.; Greuner, H.; Hurd, F.; Mendelevitch, B.; Pirsch, H.; Stadler, R.; Zangl, G.

    2011-01-01

    320 In-vessel water cooled stainless steel panels, poloidal closure plates and pumping gap panels, covering an area of approximately 100 m 2 , are used in Wendelstein7-X to protect the plasma vessel. The panels are manufactured at Deggendorf, Germany by MAN Diesel and Turbo SE. The panels consist of a laser welded sandwich of stainless steel plates together with a labyrinth of cooling channels and have a complicated geometry to fit the plasma vessel of Wendelstein 7-X. The hydraulic and mechanical stability requirements whilst maintaining the tight tolerances for the shape of the components are very demanding. The panels are designed to operate at up to an average heat load of 100 kW/m 2 and a maximum heat load of 200 kW/m 2 with a water velocity of approximately 2 m s -1 . High heat flux testing of an un-cooled panel at a time averaged load of 200 kW/m 2 for 10 s were successfully performed to support the start up phase of Wendelstein 7-X operation. Extensive testing both during manufacture and after delivery to IPP-Garching demonstrates the suitability of the delivered panels for their purpose.

  18. Plan for IER-443 Testing of the Y-12 and AWE Criticality Accident Alarm System Detectors at the Godiva IV Burst Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scorby, J. C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hickman, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hudson, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Garbett, S. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Auld, G. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Horrne, A. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Beller, T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Goda, J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Haught, C. [Y-12 National Security Complex, Oak Ridge, TN (United States); Woodrow, C. [Y-12 National Security Complex, Oak Ridge, TN (United States); Ward, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-07-24

    This document provides the scope and details of the “Plan for Testing the Y-12 and AWE Criticality Accident Alarm System Detectors at the Godiva IV Burst Reactor”. Due to the relative simplicity of the testing goals, scope, and methodology, the NCSP Manager approved execution of the test when ready. No preliminary CED-1 or final design CED-2 reports were required or issued. The test will subject Criticality Accident Alarm System (CAAS) detectors supplied by Y- 12 and AWE to very intense and short duration mixed neutron and gamma radiation fields. The goals of the test will be to (1) substantiate functionality, for both existing and newly acquired Y- 12 CAAS detectors, and (2) the ability of the AWE detectors to provide quality temporal dose information after a hypothetical criticality accident. ANSI/ANS-8.3.1997 states that the “system shall be sufficiently robust as to actuate an alarm signal when exposed to the maximum radiation expected”, which has been defined at Y-12, in Documented Safety Analyses (DSAs), to be a dose rate of 10 Rad/s. ANSI/ANS-8.3.1997 further states that “alarm actuation shall occur as a result of a minimum duration transient” which may be assumed to be 1 msec. The pulse widths and dose rates which will be achieved in this test will exceed these requirements. Pulsed radiation fields will be produced by the Godiva IV fast metal burst reactor at the National Criticality Experimental Research Center (NCERC) at the Nevada National Security Site (NNSS). The magnitude of the pulses and the relative distances to the detectors will be varied to afford a wide range of radiation fluence and pulse widths. The magnitude of the neutron and gamma fields will be determined by reactor temperature rise to fluence and dose conversions which have been previously established through extensive measurements performed under IER-147. The requirements for CAAS systems to detect and alarm under a “minimum accident of concern” as well as other

  19. Acoustic emission results obtained from testing the ZB-1 intermediate scale pressure vessel

    International Nuclear Information System (INIS)

    Hutton, P.H.; Kurtz, R.J.; Pappas, R.A.; Dawson, J.F.; Dake, L.S.; Skorpik, J.R.

    1985-09-01

    Acoustic emission (AE) monitoring of flaw growth in an intermediate scale vessel during cyclic loading at 65 0 C and 288 0 C is described in this report. The report deals with background, methodology, and results. The work discussed is of major significance in a program supported by NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. Several areas of technical concern are addressed. Results support the feasibility of effective continuous monitoring

  20. Evaluation of ductile-brittle transition behavior with neutron irradiation in nuclear reactor pressure vessel steels using small punch test

    International Nuclear Information System (INIS)

    Kim, M. C.; Lee, B. S.; Oh, Y. J.

    2003-01-01

    A Small Punch (SP) test was performed to evaluate the ductile-brittle transition temperature before and after neutron irradiation in Reactor Pressure Vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the Charpy test and Master Curve fracture toughness test in accordance with the ASTM standard E1921. The samples were taken from 1/4t location of the vessel thickness and machined into a 10x10x0.5mm dimension. Irradiation of the samples was carried out in the research reactor at KAERI (HANARO) at about 290 .deg. C of the different fluence levels respectively. SP tests were performed in the temperature range of RT to -196 .deg. C using a 2.4mm diameter ball. For the materials before and after irradiation, SP transition temperatures (T sp ), which are determined at the middle of the upper and lower SP energies, showed a linear correlation with the Charpy index temperature, T 41J . T sp from the irradiated samples was increased as the fluence level increased and was well within the deviation range of the unirradiated data. The TSP had a correlation with the reference temperature (T 0 ) from the master curve method using a pre-cracked Charpy V-notched (PCVN) specimen

  1. Behavior of underclad cracks in reactor pressure vessels - evaluation of mechanical analyses with tests on cladded mock-ups

    International Nuclear Information System (INIS)

    Moinereau, D.; Rousselier, G.; Bethmont, M.

    1993-01-01

    Innocuity of underclad flaws in the reactor pressure vessels must be demonstrated in the French safety analyses, particularly in the case of a severe transient at the end of the pressure vessel lifetime, because of the radiation embrittlement of the vessel material. Safety analyses are usually performed with elastic and elasto-plastic analyses taking into account the effect of the stainless steel cladding. EDF has started a program including experiments on large size cladded specimens and their interpretations. The purpose of this program is to evaluate the different methods of fracture analysis used in safety studies. Several specimens made of ferritic steel A508 C1 3 with stainless steel cladding, containing small artificial defects, are loaded in four-point bending. Experiments are performed at very low temperature to simulate radiation embrittlement and to obtain crack instability by cleavage fracture. Three tests have been performed on mock-ups containing a small underclad crack (with depth about 5 mn) and a fourth test has been performed on one mock-up with a larger crack (depth about 13 mn). In each case, crack instability occurred by cleavage fracture in the base metal, without crack arrest, at a temperature of about - 170 deg C. Each test is interpreted using linear elastic analysis and elastic-plastic analysis by two-dimensional finite element computations. The fracture are conservatively predicted: the stress intensity factors deduced from the computations (K cp or K j ) are always greater than the base metal toughness. The comparison between the elastic analyses (including two plasticity corrections) and the elastic-plastic analyses shows that the elastic analyses are often conservative. The beneficial effect of the cladding in the analyses is also shown : the analyses are too conservative if the cladding effects is not taken into account. (authors). 9 figs., 6 tabs., 10 refs

  2. The Drift Burst Hypothesis

    OpenAIRE

    Christensen, Kim; Oomen, Roel; Renò, Roberto

    2016-01-01

    The Drift Burst Hypothesis postulates the existence of short-lived locally explosive trends in the price paths of financial assets. The recent US equity and Treasury flash crashes can be viewed as two high profile manifestations of such dynamics, but we argue that drift bursts of varying magnitude are an expected and regular occurrence in financial markets that can arise through established mechanisms such as feedback trading. At a theoretical level, we show how to build drift bursts into the...

  3. Tests on model of a prestressed concrete nuclear pressure vessel with multiple cavities

    International Nuclear Information System (INIS)

    Favre, R.; Koprna, M.; Jaccoud, J.P.

    1977-01-01

    The prestressed concrete pressure vessel (prototype) is a cylinder having a diameter of 48 m and a height of 39 m. It has 25 vertical cavities (reactor, heat exchangers, heat recuperators) and 3 horizontal cavities (gas turbines of 500 kw). The cavities are closed by plugs, and their tightness is ensured by a steel lining. A model, on a scale of 1/20, made of microconcrete, was loaded in several cycles, by a uniform inner pressure in the cavities, increasing to the point of failure. The three successive stages were examined: stage of globally elastic behavior, cracking stage, ultimate stage. The behavior of the model is globally elastic up to an inner pressure of 120 to 130 kp/cm 2 , corresponding to about twice the maximum pressure of service, equal to 65 kp/cm 2 . The prestressed tendons at this stage show practically no stress increase. The first detectable cracks appear on the lateral side half-way up the model, as soon as the pressure exceeded 120 kp/cm 2 . From 150-165 kp/cm 2 , the cracking stage can be considered as achieved and the main crack pattern entirely formed. A horizontal crack continues in the middle of the barrel, as well as vertical cracks at each outer cavity. Beyond a pressure of 150-165 kp/cm 2 the ultimate stage begins. The strains of the stresses in the tendons grow more rapidly. The steel lining is highly solicited. Above about 210 kp/cm 2 the model behaves like a structure composed of a group of concrete blocks bound by the tendons and the lining. The failure (240 kp/cm 2 ) occurred through a mechanism of ejection and bending of the concrete ring at the periphery of the barrel of the vessel, which was solicited mainly in tension

  4. Burst protected nuclear reactor plant with PWR

    International Nuclear Information System (INIS)

    Harand, E.; Michel, E.

    1978-01-01

    In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

  5. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  6. Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention. Pt. II. Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

    International Nuclear Information System (INIS)

    Chu, T.Y.; Bentz, J.H.; Slezak, S.E.; Pasedag, W.F.

    1997-01-01

    For pt.I see ibid., p.77-88 (1997). This paper summarizes the results of a reactor-scale ex-vessel boiling experiment for assessing the flooded cavity design of the heavy water new production reactor. The simulated reactor vessel has a cylindrical diameter of 3.7 m and a torispherical bottom head. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling mainly results from the gravity head, which in turn results from flooding the side of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid-solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion. The results show that, under prototypic heat load and heat flux distributions, the flooded cavity will be effective for in-vessel core retention in the heavy water new production reactor. The results also demonstrate that the heat dissipation requirement for in-vessel core retention, for the central region of the lower head of an AP-600 advanced light water reactor, can be met with the flooded cavity design. (orig.)

  7. Specific Fluorescence in Situ Hybridization (FISH) Test to Highlight Colonization of Xylem Vessels by Xylella fastidiosa in Naturally Infected Olive Trees (Olea europaea L.)

    Science.gov (United States)

    Cardinale, Massimiliano; Luvisi, Andrea; Meyer, Joana B.; Sabella, Erika; De Bellis, Luigi; Cruz, Albert C.; Ampatzidis, Yiannis; Cherubini, Paolo

    2018-01-01

    The colonization behavior of the Xylella fastidiosa strain CoDiRO, the causal agent of olive quick decline syndrome (OQDS), within the xylem of Olea europaea L. is still quite controversial. As previous literature suggests, even if xylem vessel occlusions in naturally infected olive plants were observed, cell aggregation in the formation of occlusions had a minimal role. This observation left some open questions about the whole behavior of the CoDiRO strain and its actual role in OQDS pathogenesis. In order to evaluate the extent of bacterial infection in olive trees and the role of bacterial aggregates in vessel occlusions, we tested a specific fluorescence in situ hybridization (FISH) probe (KO 210) for X. fastidiosa and quantified the level of infection and vessel occlusion in both petioles and branches of naturally infected and non-infected olive trees. All symptomatic petioles showed colonization by X. fastidiosa, especially in the larger innermost vessels. In several cases, the vessels appeared completely occluded by a biofilm containing bacterial cells and extracellular matrix and the frequent colonization of adjacent vessels suggested a horizontal movement of the bacteria. Infected symptomatic trees had 21.6 ± 10.7% of petiole vessels colonized by the pathogen, indicating an irregular distribution in olive tree xylem. Thus, our observations point out the primary role of the pathogen in olive vessel occlusions. Furthermore, our findings indicate that the KO 210 FISH probe is suitable for the specific detection of X. fastidiosa. PMID:29681910

  8. Intermittent Theta Burst Over M1 May Increase Peak Power of a Wingate Anaerobic Test and Prevent the Reduction of Voluntary Activation Measured with Transcranial Magnetic Stimulation.

    Science.gov (United States)

    Giboin, Louis-Solal; Thumm, Patrick; Bertschinger, Raphael; Gruber, Markus

    2016-01-01

    Despite the potential of repetitive transcranial magnetic stimulation (rTMS) to improve performances in patients suffering from motor neuronal afflictions, its effect on motor performance enhancement in healthy subjects during a specific sport task is still unknown. We hypothesized that after an intermittent theta burst (iTBS) treatment, performance during the Wingate Anaerobic Test (WAnT) will increase and supraspinal fatigue following the exercise will be lower in comparison to a control treatment. Ten subjects participated in two randomized experiments consisting of a WAnT 5 min after either an iTBS or a control treatment. We determined voluntary activation (VA) of the right knee extensors with TMS (VATMS) and with peripheral nerve stimulation (VAPNS) of the femoral nerve, before and after the WAnT. T-tests were applied to the WAnT results and a two way within subject ANOVA was applied to VA results. The iTBS treatment increased the peak power and the maximum pedalling cadence and suppressed the reduction of VATMS following the WAnT compared to the control treatment. No behavioral changes related to fatigue (mean power and fatigue index) were observed. These results indicate for the first time that iTBS could be used as a potential intervention to improve anaerobic performance in a sport specific task.

  9. Intermittent theta burst over M1 may increase peak power of a Wingate anaerobic test and prevent the reduction of voluntary activation measured with transcranial magnetic stimulation

    Directory of Open Access Journals (Sweden)

    Louis-Solal Giboin

    2016-07-01

    Full Text Available Despite the potential of repetitive transcranial magnetic stimulation (rTMS to improve performances in patients suffering from motor neuronal afflictions, its effect on motor performance enhancement in healthy subjects during a specific sport task is still unknown. We hypothesised that after an intermittent theta burst (iTBS treatment, performance during the Wingate Anaerobic Test (WAnT, will increase and supraspinal fatigue following the exercise will be lower in comparison to a control treatment.Ten subjects participated in two randomised experiments consisting of a WAnT 5 minutes after either an iTBS or a control treatment. We determined voluntary activation (VA of the right knee extensors with TMS (VATMS and with peripheral nerve stimulation (VAPNS of the femoral nerve, before and after the WAnT. T-tests were applied to the WAnT results and a 2 way within subject ANOVA was applied to VA results. The iTBS treatment increased the peak power and the maximum pedalling cadence and suppressed the reduction of VATMS following the WAnT compared to the control treatment. No behavioural changes related to fatigue (mean power and fatigue index were observed.These results indicate for the first time that iTBS could be used as a potential intervention to improve anaerobic performance in a sport specific task.

  10. Composite Overwrap Pressure Vessels: Mechanics and Stress Rupture Lifting Philosophy

    Science.gov (United States)

    Thesken, John C.; Murthy, Pappu L. N.; Phoenix, S. L.

    2009-01-01

    The NASA Engineering and Safety Center (NESC) has been conducting an independent technical assessment to address safety concerns related to the known stress rupture failure mode of filament wound pressure vessels in use on Shuttle and the International Space Station. The Shuttle s Kevlar-49 (DuPont) fiber overwrapped tanks are of particular concern due to their long usage and the poorly understood stress rupture process in Kevlar-49 filaments. Existing long term data show that the rupture process is a function of stress, temperature and time. However due to the presence of load sharing liners and the complex manufacturing procedures, the state of actual fiber stress in flight hardware and test articles is not clearly known. Indeed nonconservative life predictions have been made where stress rupture data and lifing procedures have ignored the contribution of the liner in favor of applied pressure as the controlling load parameter. With the aid of analytical and finite element results, this paper examines the fundamental mechanical response of composite overwrapped pressure vessels including the influence of elastic plastic liners and degraded/creeping overwrap properties. Graphical methods are presented describing the non-linear relationship of applied pressure to Kevlar-49 fiber stress/strain during manufacturing, operations and burst loadings. These are applied to experimental measurements made on a variety of vessel systems to demonstrate the correct calibration of fiber stress as a function of pressure. Applying this analysis to the actual qualification burst data for Shuttle flight hardware revealed that the nominal fiber stress at burst was in some cases 23 percent lower than what had previously been used to predict stress rupture life. These results motivate a detailed discussion of the appropriate stress rupture lifing philosophy for COPVs including the correct transference of stress rupture life data between dissimilar vessels and test articles.

  11. Composite Overwrap Pressure Vessels: Mechanics and Stress Rupture Lifing Philosophy

    Science.gov (United States)

    Thesken, John C.; Murthy, Pappu L. N.; Phoenix, Leigh

    2007-01-01

    The NASA Engineering and Safety Center (NESC) has been conducting an independent technical assessment to address safety concerns related to the known stress rupture failure mode of filament wound pressure vessels in use on Shuttle and the International Space Station. The Shuttle's Kevlar-49 fiber overwrapped tanks are of particular concern due to their long usage and the poorly understood stress rupture process in Kevlar-49 filaments. Existing long term data show that the rupture process is a function of stress, temperature and time. However due to the presence of load sharing liners and the complex manufacturing procedures, the state of actual fiber stress in flight hardware and test articles is not clearly known. Indeed non-conservative life predictions have been made where stress rupture data and lifing procedures have ignored the contribution of the liner in favor of applied pressure as the controlling load parameter. With the aid of analytical and finite element results, this paper examines the fundamental mechanical response of composite overwrapped pressure vessels including the influence of elastic-plastic liners and degraded/creeping overwrap properties. Graphical methods are presented describing the non-linear relationship of applied pressure to Kevlar-49 fiber stress/strain during manufacturing, operations and burst loadings. These are applied to experimental measurements made on a variety of vessel systems to demonstrate the correct calibration of fiber stress as a function of pressure. Applying this analysis to the actual qualification burst data for Shuttle flight hardware revealed that the nominal fiber stress at burst was in some cases 23% lower than what had previously been used to predict stress rupture life. These results motivate a detailed discussion of the appropriate stress rupture lifing philosophy for COPVs including the correct transference of stress rupture life data between dissimilar vessels and test articles.

  12. Application of computer techniques to charpy impact testing of irradiated pressure vessel steels

    International Nuclear Information System (INIS)

    Landow, M.P.; Fromm, E.O.; Perrin, J.S.

    1982-01-01

    A Rockwell AIM 65 microcomputer has been modified to control a remote Charpy V-notch impact test machine. It controls not only handling and testing of the specimen but also transference and storage of instrumented Charpy test data. A system of electrical solenoid activated pneumatic cylinders and switches provides the interface between the computer and the test apparatus. A command language has been designated that allows the operator to command checkout, test procedure, and data storage via the computer. Automatic compliance with ASTM test procedures is built into the program

  13. Gamma Ray Bursts - Observations

    Science.gov (United States)

    Gehrels, N.; Cannizzo, J. K.

    2010-01-01

    We are in an exciting period of discovery for gamma-ray bursts. The Swift observatory is detecting 100 bursts per year, providing arcsecond localizations and sensitive observations of the prompt and afterglow emission. The Fermi observatory is observing 250 bursts per year with its medium-energy GRB instrument and about 10 bursts per year with its high-energy LAT instrument. In addition, rapid-response telescopes on the ground are providing new capabilities to study optical emission during the prompt phase and spectral signatures of the host galaxies. The combined data set is enabling great advances in our understanding of GRBs including afterglow physics, short burst origin, and high energy emission.

  14. Behaviour of the nozzle corner region during the first phase of the fatigue test on scaled models of pressure vessels (JRC Vessel A)

    International Nuclear Information System (INIS)

    Jovanovic, A.; Lucia, A.C.; Brunnhuber, R.; Elbaz, J.M.; Schwarz, U.

    1987-01-01

    The work presented here deals mainly with the stress and fracture mechanics aspects of the first phase of the structural reliability experimental program based on the scaled experimental vessel at Ispra. The overall research and experiments make also part of the structural reliability assessment extrapolation pattern for the full-scale structures. The work presented here deals with the problem of the nozzle corner cracks induced by the fatigue under the pressure cycling

  15. Reduction of the number of defect signals in pressure vessel welds by a phased array ultrasonic test technology qualified beforehand in a blind test according to PDI specifications

    International Nuclear Information System (INIS)

    Mohr, F.

    2007-01-01

    In German-language countries, ultrasonic testing of reactor pressure vessel welds in the context of recurrent inspection is based on the KTA rules. This test philosophy is based on the recording of all data of a test section and repeated comparison of these data at regular intervals. Each and every change during operation is displayed. There are many components in which no changes are observed over longer periods of time. Optimisation of the test procedure and test periods requires accurate knowledge of the component condition. This necessitates accurate data of available defects. However, current techniques only provide data for comparative analysis on the basis of reflectivity. Data on the length and depth of a relevant defect can only be obtained by qualified sizing techniques. The PDI programme provides exact rules for qualification of techniques for a given application. Using a PDI qualification with personal blind tests for all data evaluators, one obtains a basis for accurate defect dimensioning and thus for optimisation. In cooperation with KKL, IntelligeNDT AREVA in 2006 successfully underwent the PDI qualification process for phased array testing of longitudinal and circumferential welds in reactor pressure vessels. In addition to this qualification, a comparison was made with the results of the conventionally applied, KTA-oriented test procedure. One of the key elements of qualification is the characterisation of defects, i.e. the distinction between relevant and non-relevant data, which will help to reduce the displayed data. The contribution presents the results and experience of the qualification as well as a comparison of standard testing with a tandem function with the results of phased array testing. (orig.)

  16. Nuclear Power Plant Prestressed Concrete Containment Vessel Structure Monitoring during Integrated Leakage Rate Testing Using Fiber Bragg Grating Sensors

    Directory of Open Access Journals (Sweden)

    Jinke Li

    2017-04-01

    Full Text Available As the last barrier of nuclear reactor, prestressed concrete containment vessels (PCCVs play an important role in nuclear power plants (NPPs. To test the mechanical property of PCCV during the integrated leakage rate testing (ILRT, a fiber Bragg grating (FBG sensor was used to monitor concrete strain. In addition, a finite element method (FEM model was built to simulate the progress of the ILRT. The results showed that the strain monitored by FBG had the same trend compared to the inner pressure variation. The calculation results showed a similar trend compared with the monitoring results and provided much information about the locations in which the strain sensors should be installed. Therefore, it is confirmed that FBG sensors and FEM simulation are very useful in PCCV structure monitoring.

  17. Review of design analysis and site installation method of ASME vessel of fuel test loop of Hanaro

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Sung Won; Kim, Jun Yeon

    1997-02-01

    The major goal of this report is to take the advantages under control, and provide basic solutions for field installation. In order to access to technical availability, the scope, limitation, and possibilities of design requirements are carefully considered in this. The chapter 3 deals with seismic stress analysis of vessels, using manufacturers` finite element analysis data. The test of manufactured equipment is scheduled to hold near future. The evaluation criteria and inspection specifications for the quality release are well illustrated in the chapter IV to efficiently check out the items. The full considerations for field installation are also reviewed with that in terms of space limit. Following the inspection and test of manufacturer`s shop, the results will be promptly reported. Based on installation principles and the analysis in the report, updated procedure and methodology for the installation will be applied to the field in more details and better breakdown precision in the coming years. (author). 14 refs., 23 tabs., 26 figs.

  18. Review of design analysis and site installation method of ASME vessel of fuel test loop of Hanaro

    International Nuclear Information System (INIS)

    Cho, Sung Won; Kim, Jun Yeon.

    1997-02-01

    The major goal of this report is to take the advantages under control, and provide basic solutions for field installation. In order to access to technical availability, the scope, limitation, and possibilities of design requirements are carefully considered in this. The chapter 3 deals with seismic stress analysis of vessels, using manufacturers' finite element analysis data. The test of manufactured equipment is scheduled to hold near future. The evaluation criteria and inspection specifications for the quality release are well illustrated in the chapter IV to efficiently check out the items. The full considerations for field installation are also reviewed with that in terms of space limit. Following the inspection and test of manufacturer's shop, the results will be promptly reported. Based on installation principles and the analysis in the report, updated procedure and methodology for the installation will be applied to the field in more details and better breakdown precision in the coming years. (author). 14 refs., 23 tabs., 26 figs

  19. Post-test calculations of out-of-pile single rod burst experiments with Argentine ZRY-4 tubes using SSYST codes system

    International Nuclear Information System (INIS)

    Nassini, H.; Meyder, R.

    1988-11-01

    The calculations were performed with SSYST-3 code system. The experiments have been carried out with shortened REBEKA simulators. The main goal was to verify whether the model for describing the Zircaloy cladding deformation and rupture at high temperatures (NORA2 model) is able to predict the Argentine cladding material behaviour under such conditions. A good agreement between experimental results and model predictions was found in the range of temperatures from 700 to 850 0 C (in α phase and first half of α+β transition phase). There were some discrepances at higher temperatures, probably due to the strong influence of heating rate on material properties. According to the results of these post-test calculations, it can be concluded that NORA2 model is able to well describe the Argentine Zircaloy cladding deformation and rupture under LOCA conditions, because there was a good agreement in the range of burst (ballooning) temperatures which are expected in this type of accident. (orig./HP) [de

  20. The composition of aerosols generated during a severe reactor accident: Experimental results from the Power Burst Facility Severe Fuel Damage Test 1-4

    International Nuclear Information System (INIS)

    Petti, D.A.; Hobbins, R.R.; Hagrman, D.L.

    1994-01-01

    Experimental results on fission product and aerosol release during the Power Burst Facility Severe Fuel Damages (SFD) Test 1-4 are examined to determine the composition of aerosols that would be generated during a severe reactor accident. The SFD 1-4 measured aerosol contained significant quantities of volatile fission products (VFPs) (cesium, iodine, tellurium), control materials (silver and cadmium), and structural materials (tin), indicating that fission product release, vaporization of control material, and release of tin from oxidized Zircaloy were all important aerosol sources. On average the aerosol composition is between one-quarter and one-half VFPs (especially cesium), with the remainder being control material (especially cadmium), and structural material (especially tin). Source term computer codes like CORSOR-M tend to overpredict the release of structural and control rod material relative to fission products by a factor of between 2 and 15 because the models do not account for relocation of molten control, fuel, and structural material during the degradation process, which tends to reduce the aerosol source. The results indicate that the aerosol generation in a severe reactor accident is intimately linked to the core degradation process. They recommend that these results be used to improve the models in source term computer codes

  1. Endurance test report of rubber sealing materials for the containment vessel

    International Nuclear Information System (INIS)

    Yamamoto, R.; Watanabe, K.; Hanashima, K.

    2015-01-01

    In the event of a nuclear power plant accident such as a core meltdown and a cooling system failure, the containment contains radioactive materials released from the reactor pressure vessel to reduce the activity of the radioactive materials and the effects of radiation in the vicinity of the plant. Since high sealing performance and high pressure resistance are required of the containment, a silicone or EPDM rubber gasket with high heat and radiation resistance is used for the sealing of the sealing boundary of the containment. In recent years, it has been shown that a large amount of steam is released into the containment in the case of a severe accident. Consequently, radiation resistance at high temperature as well as steam resistance is required of the rubber gasket placed at the sealing boundary. However, the steam resistance of silicone rubber is not necessarily as good as that of EPDM rubber. Therefore, it is necessary to evaluate the sealing characteristics of rubber gaskets in such a degrading environment in a severe accident. O. Kato et al. [1] conducted a study on the degradation status of rubber gaskets and their application limits at high temperature. However, few studies have evaluated rubber gaskets in high-temperature radiation and steam environments. In this study, we degraded silicone rubber and EPDM rubber used for the containment in the high-temperature radiation and steam environments expected to occur in a severe accident and evaluated the useful life of the rubber as a sealing material by estimating the change in its performance as a sealing material from the change in permanent compressive strain in the rubber. (author)

  2. Compilation of three-dimensional coordinates and specific data of the instrumentation of the prestressed concrete pressure vessel/high temperature helium test rig

    International Nuclear Information System (INIS)

    Klausinger, D.

    1977-04-01

    The positions of the thermoelements, strain gauges of various types, and of Gloetzl instruments installed by SGAE in the model vessel of the Common Project Prestressed Concrete Pressure Vessel/High Temperature Helium Test Rig are defined in cylindrical coordinates. The specific data of the instruments are given like configuration of multiple instruments; type, group and number of the instrument; number of cable and of channel; calibration factors; resistances of instruments and cables. (author)

  3. Materials interaction tests to identify base and coating materials for an enhanced in-vessel core catcher design

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Knudson, D.L.; Condie, K.G.; Swank, W.D. [Idaho National Engineering and Environmental Laboratory, Idaho Falls ID (United States); Cheung, F.B. [Pennsylvania State University, Department of Mechanical and Nuclear Engineering, University Park PA (United States); Suh, K.Y. [Seoul National University, Department of Nuclear Engineering, Seoul (Korea, Republic of); Kim, S.B. [Korea Atomic Energy Research Institute, Severe Accident Research Project, Taejon (Korea, Republic of)

    2004-07-01

    An enhanced in-vessel core catcher is being designed and evaluated, it must ensure In-Vessel Retention of core materials that may relocate under severe accident conditions in advanced reactors. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. If needed, the core catcher can be manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary): a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an insulating oxide coating material on top of the base material, which resists interactions with high-temperature core materials; and an optional coating on the bottom side of the base material to prevent any potential oxidation of the base material during the lifetime of the reactor. Initial evaluations suggest that a thermally-sprayed oxide material is the most promising candidate insulator coating for a core catcher. Tests suggest that 2 coatings can provide adequate protection to a stainless steel core catcher: -) a 500 {mu}m thick zirconium dioxide coating over a 100-200 {mu}m Inconel 718 bond coating, and -) a 500 {mu}m thick magnesium zirconate coating.

  4. FFTF and CRBRP reactor vessels

    International Nuclear Information System (INIS)

    Morgan, R.E.

    1977-01-01

    The Fast Flux Test Facility (FFTF) reactor vessel and the Clinch River Breeder Reactor Plant (CRBRP) reactor vessel each serve to enclose a fast spectrum reactor core, contain the sodium coolant, and provide support and positioning for the closure head and internal structure. Each vessel is located in its reactor cavity and is protected by a guard vessel which would ensure continued decay heat removal capability should a major system leak develop. Although the two plants have significantly different thermal power ratings, 400 megawatts for FFTF and 975 megawatts for CRBRP, the two reactor vessels are comparable in size, the CRBRP vessel being approximately 28% longer than the FFTF vessel. The FFTF vessel diameter was controlled by the space required for the three individual In-Vessel Handling Machines and Instrument Trees. Utilization of the triple rotating plug scheme for CRBRP refueling enables packaging of the larger CRBRP core in a vessel the same diameter as the FFTF vessel

  5. An Ancient Transcription Factor Initiates the Burst of piRNA Production During Early Meiosis in Mouse Testes

    Science.gov (United States)

    Li, Xin Zhiguo; Roy, Christian K.; Dong, Xianjun; Bolcun-Filas, Ewelina; Wang, Jie; Han, Bo W.; Xu, Jia; Moore, Melissa J.; Schimenti, John C.; Weng, Zhiping; Zamore, Phillip D.

    2013-01-01

    SUMMARY Animal germ cells produce PIWI-interacting RNAs (piRNAs), small silencing RNAs that suppress transposons and enable gamete maturation. Mammalian transposon-silencing piRNAs accumulate early in spermatogenesis, whereas pachytene piRNAs are produced later during post-natal spermatogenesis and account for >95% of all piRNAs in the adult mouse testis. Mutants defective for pachytene piRNA pathway proteins fail to produce mature sperm, but neither the piRNA precursor transcripts nor the trigger for pachytene piRNA production is known. Here, we show that the transcription factor A-MYB initiates pachytene piRNA production. A-MYB drives transcription of both pachytene piRNA precursor RNAs and the mRNAs for core piRNA biogenesis factors, including MIWI, the protein through which pachytene piRNAs function. A-MYB regulation of piRNA pathway proteins and piRNA genes creates a coherent feed-forward loop that ensures the robust accumulation of pachytene piRNAs. This regulatory circuit, which can be detected in rooster testes, likely predates the divergence of birds and mammals. PMID:23523368

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

    International Nuclear Information System (INIS)

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

    1995-07-01

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

  7. Medical Tests and Procedures for Finding and Treating Heart and Blood Vessel Disease

    Science.gov (United States)

    ... Coronary calcium scan Health care providers use electron-beam computed tomography (to-MOG-rah-fee) (EBCT) or ... how the heart performs during activity, such as walking on a moving treadmill. A medication stress test ...

  8. Nondestructive testing during the fabrication of pressure vessels with half pipe jackets

    International Nuclear Information System (INIS)

    Scherner, D.

    1985-01-01

    The most important precondition to guarantee the optimum quality of half pipe jackets is the precise fixing and observance of the manufacturing conditions. For this reason the manufacturing conditions are explained in detail. The second important point is the test for gas tightness of the half pipe jacket system. The sources of mistakes in connection with the test for gas tightness are of fundamental importance. (orig./PW) [de

  9. Advanced communications technology satellite high burst rate link evaluation terminal power control and rain fade software test plan, version 1.0

    Science.gov (United States)

    Reinhart, Richard C.

    1993-01-01

    The Power Control and Rain Fade Software was developed at the NASA Lewis Research Center to support the Advanced Communications Technology Satellite High Burst Rate Link Evaluation Terminal (ACTS HBR-LET). The HBR-LET is an experimenters terminal to communicate with the ACTS for various experiments by government, university, and industry agencies. The Power Control and Rain Fade Software is one segment of the Control and Performance Monitor (C&PM) Software system of the HBR-LET. The Power Control and Rain Fade Software automatically controls the LET uplink power to compensate for signal fades. Besides power augmentation, the C&PM Software system is also responsible for instrument control during HBR-LET experiments, control of the Intermediate Frequency Switch Matrix on board the ACTS to yield a desired path through the spacecraft payload, and data display. The Power Control and Rain Fade Software User's Guide, Version 1.0 outlines the commands and procedures to install and operate the Power Control and Rain Fade Software. The Power Control and Rain Fade Software Maintenance Manual, Version 1.0 is a programmer's guide to the Power Control and Rain Fade Software. This manual details the current implementation of the software from a technical perspective. Included is an overview of the Power Control and Rain Fade Software, computer algorithms, format representations, and computer hardware configuration. The Power Control and Rain Fade Test Plan provides a step-by-step procedure to verify the operation of the software using a predetermined signal fade event. The Test Plan also provides a means to demonstrate the capability of the software.

  10. Structural integrity test of prestressed concrete containment vessel for Tsuruga Unit No. 2 Nuclear Power Station

    International Nuclear Information System (INIS)

    Tamura, S.; Nagata, K.; Takeda, T.; Yamaguchi, T.; Nakayama, T.

    1987-01-01

    In introducing the PCCV to Japan, various verification tests were carried out to understand the structural performance of the PCCV and confirm the reliability of its design. In addition to those tests, a Structural Integrity Test (SIT) was conducted in Feb. 1986 as a final acceptance test. This report discusses the results of the SIT on the PCCV. The test was carried out simultaneously with an Integrated Leak Rate Test (ILRT) under the same pressure sequence. 1) Pressure-displacement relationships and pressure-strain relationships were more or less linear. 2) The measured displacement values at the maximum pressure (4.5 kgf/cm 2 G) corresponded well with calculated values. Correspondence with converted displacement obtained from strain and measured displacement was also good. 3) The residual displacement when 24 hours had elapsed after completion of depressurization was not more than 10% of the displacement at the maximum pressure. 4) The variation in tendon force at the maximum pressure is smaller than the calculated value in proportion to the elongation of the PCCV. 5) Although fine surface cracks due to shrinkage of concrete were seen, new structural cracks due to pressure were not observed. The leakage rate was evaluated at 0.016% of volume per day. It is much smaller than the design value of 0.1% of volume per day. (orig./HP)

  11. Structural design of shield-integrated thin-wall vacuum vessel and manufacturing qualification tests for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Shimizu, Katsusuke; Shibui, Masanao; Koizumi, Koichi; Kanamori, Naokazu; Nishio, Satoshi; Sasaki, Takashi; Tada, Eisuke

    1992-09-01

    Conceptual design of shield-integrated thin-wall vacuum vessel has been done for ITER (International Thermonuclear Experimental Reactor). The vacuum vessel concept is based on a thin-double-wall structure, which consists of inner and outer plates and rib stiffeners. Internal shielding structures, which provide neutron irradiation shielding to protect TF coils, are set up between the inner plate and the outer plate of the vessel to avoid complexity of machine systems such as supporting systems of blanket modules. The vacuum vessel is assembled/disassembled by remote handling, so that welding joints are chosen as on-site joint method from reliability of mechanical strength. From a view point of assembling TF coils, the vacuum vessel is separated at the side of port, and is divided into 32 segments similar to the ITER-CDA reference design. Separatrix sweeping coils are located in the vacuum vessel to reduce heat fluxes onto divertor plates. Here, the coil structure and attachment to the vacuum vessel have been investigated. A sectorized saddle-loop coil is available for assembling and disassembling the coil. To support electromagnetic loads on the coils, they are attached to the groove in the vacuum vessel by welding. Flexible multi-plate supporting structure (compression-type gravity support), which was designed during CDA, is optimized by investigating buckling and frequency response properties, and concept on manufacturing and fabrication of the gravity support are proposed. Partial model of the vacuum vessel is manufactured for trial, so that fundamental data on welding and fabrication are obtained. From mechanical property tests of weldment and partial models, mechanical intensity and behaviors of the weldment are obtained. Informations on FEM-modeling are obtained by comparing analysis results with experimental results. (author)

  12. Possible galactic origin of. gamma. -ray bursts

    Energy Technology Data Exchange (ETDEWEB)

    Manchanda, R K; Ramsden, D [Southampton Univ. (UK). Dept. of Physics

    1977-03-31

    It is stated that extragalactic models for the origin of non-solar ..gamma..-ray bursts include supernova bursts in remote galaxies, and the collapse of the cores of active stars, whilst galactic models are based on flare stars, thermonuclear explosions in neutron stars and the sudden accretion of cometary gas on to neutron stars. The acceptability of any of these models may be tested by the observed size spectrum of the ..gamma..-ray bursts. The extragalactic models predict a power law spectrum with number index -1.5, whilst for the galactic models the number index will be -1. Experimental data on ..gamma..-ray bursts is, however, still meagre, and so far only 44 confirmed events have been recorded by satellite-borne instruments. The number spectrum of the observed ..gamma..-ray bursts indicates that the observed distribution for events with an energy < 10/sup -4/ erg/cm/sup 2/ is flat; this makes the choice of any model completely arbitrary. An analysis of the observed ..gamma..-ray events is here presented that suggests very interesting possibilities for their origin. There appears to be a preferred mean energy for ..gamma..-ray bursts; some 90% of the recorded events show a mean energy between 5 x 10/sup -5/ and 5 x 10/sup -4/ erg/cm/sup 2/, contrary to the predicted characteristics of the number spectrum of various models. A remarkable similarity is found between the distribution of ..gamma..-ray bursts and that of supernova remnants, suggesting a genetic relationship between the two and the galactic origin of the ..gamma..-ray bursts, and the burst source could be identified with completely run down neutron stars, formed during supernova explosions.

  13. Research vessels

    Digital Repository Service at National Institute of Oceanography (India)

    Rao, P.S.

    The role of the research vessels as a tool for marine research and exploration is very important. Technical requirements of a suitable vessel and the laboratories needed on board are discussed. The history and the research work carried out...

  14. Analyses and tests for the baking system of the RFX vacuum vessel by eddy currents

    International Nuclear Information System (INIS)

    Collarin, P.; Sonato, P.; Zaccaria, P.; Zollino, G.

    1995-01-01

    The electrical, thermal and mechanical analyses carried out for the design of a new baking system for RFX by eddy currents are presented. The results of an experimental test on RFX with low heating power are reported as well. They gave confidence in the numerical analyses so as the working conditions with the nominal heating power were computed. (orig.)

  15. Analyses and tests for the baking system of the RFX vacuum vessel by eddy currents

    Energy Technology Data Exchange (ETDEWEB)

    Collarin, P. [Gruppo di Padova per Ricerche sulla Fusione, Univ. di Padova (Italy); Sonato, P. [Gruppo di Padova per Ricerche sulla Fusione, Univ. di Padova (Italy); Zaccaria, P. [Gruppo di Padova per Ricerche sulla Fusione, Univ. di Padova (Italy); Zollino, G. [Gruppo di Padova per Ricerche sulla Fusione, Univ. di Padova (Italy)

    1995-12-31

    The electrical, thermal and mechanical analyses carried out for the design of a new baking system for RFX by eddy currents are presented. The results of an experimental test on RFX with low heating power are reported as well. They gave confidence in the numerical analyses so as the working conditions with the nominal heating power were computed. (orig.).

  16. Development and testing of bumper limiter of aluminum alloy vacuum vessel for reacting plasma experiment

    International Nuclear Information System (INIS)

    Uchikawa, T.; Fujiwara, M.; Ioki, K.; Irie, T.; Nayama, R.; Nishikawa, M.; Onozuka, M.; Tomita, M.

    1985-01-01

    Two types of graphite bumper limiters were designed and trially fabricated for a reacting plasma device, R-tokamak. High heat load tests were conducted to examine thermal behavior and thermal shock resistance of the limiters by using a 100kW electron beam facility. The experimental data were compared with the results of 3-D thermal analysis

  17. Development and testing of bumper limiter of aluminum alloy vacuum vessel for reacting plasma experiment

    Energy Technology Data Exchange (ETDEWEB)

    Uchikawa, T.; Fujiwara, M.; Ioki, K.; Irie, T.; Nayama, R.; Nishikawa, M.; Onozuka, M.; Tomita, M.

    1985-07-01

    Two types of graphite bumper limiters were designed and trially fabricated for a reacting plasma device, R-tokamak. High heat load tests were conducted to examine thermal behavior and thermal shock resistance of the limiters by using a 100kW electron beam facility. The experimental data were compared with the results of 3-D thermal analysis.

  18. Application of a microcomputer-based data system to ultimate strength tests of pressure vessels

    International Nuclear Information System (INIS)

    Woodfin, R.L.

    1983-01-01

    Sandia is evaluating the degree to which existing finite element computer codes can predict the structural behavior of containment buildings beyond the elastic range. The study also is attempting to discover whether any intrinsic functional relationship exists relating leakage to pressure. Test equipment and instrumentation are described

  19. Ultrasonic data acquisition installation for basis and in-service testing of nuclear pressure vessels

    International Nuclear Information System (INIS)

    Gutmann, G.; Engl, G.

    1976-01-01

    The safety of nuclear installations requires continuous safety inspections during construction and operation. Essential parts of this safety inspection are the basis and in-line inspections. For this purpose installation systems are used which allow an optimal statement to be made regarding the conditions of tested components

  20. Swift pointing and gravitational-wave bursts from gamma-ray burst events

    International Nuclear Information System (INIS)

    Sutton, Patrick J; Finn, Lee Samuel; Krishnan, Badri

    2003-01-01

    The currently accepted model for gamma-ray burst phenomena involves the violent formation of a rapidly rotating solar-mass black hole. Gravitational waves should be associated with the black-hole formation, and their detection would permit this model to be tested. Even upper limits on the gravitational-wave strength associated with gamma-ray bursts could constrain the gamma-ray burst model. This requires joint observations of gamma-ray burst events with gravitational and gamma-ray detectors. Here we examine how the quality of an upper limit on the gravitational-wave strength associated with gamma-ray bursts depends on the relative orientation of the gamma-ray-burst and gravitational-wave detectors, and apply our results to the particular case of the Swift Burst-Alert Telescope (BAT) and the LIGO gravitational-wave detectors. A result of this investigation is a science-based 'figure of merit' that can be used, together with other mission constraints, to optimize the pointing of the Swift telescope for the detection of gravitational waves associated with gamma-ray bursts

  1. Improved ultrasonic nondestructive testing of pressure vessels. Annual progress report, August 1, 1975--July 31, 1976

    International Nuclear Information System (INIS)

    Frederick, J.R.; Fairchild, R.C.; Anderson, B.H.

    1977-07-01

    A synthetic aperture focusing technique for ultrasonic testing (SAFT UT) is described. The technique employs a single scanned transducer operating in pulse-echo mode with digital data acquisition and synthetic aperture post-processing to provide high lateral and longitudinal resolution. The extension of previously developed algorithms to provide volumetric processing and display is described. The design of a refreshed grey-scale display to provide interactive display of SAFT UT data is described

  2. Ultrasonic testing of large blocks for prestressed cast iron pressure vessels

    International Nuclear Information System (INIS)

    Stelling, H.A.

    1979-01-01

    Ultrasonic tests were made on plate specimen and large blocks of perlit cast iron with lamellar graphite. Aims of the investigations were the control of material porperties, the flaw detection and flaw classification. The material properties were classified by sound velocity and attenuation measurements. Flaw detection and flaw size estimation methods were modified with regard to the acoustic properties, the microstructure and the reflectivity of typical flaws in castings. Special localisation and flaw size estimation techniques are discussed. (orig.)

  3. Experimental tests on buckling of ellipsoidal vessel heads under internal pressure

    International Nuclear Information System (INIS)

    Alix, Michel; Roche, Roland.

    1979-01-01

    Seventeen heads made out of metal sheets -by cold working- were tested. Three different metals were used - carbon steel, austenitic steel, and aluminium alloy. Nominal dimensions were: diameter D 500 mm height H 50 and 100 mm thickness to diameter ratio t/D in the range 0.001-0.005. The heads had a good axisymmetric shape, but that the thickness was varying along the ellipse. Material characteristic of each head was given by a tensile test (strain-stress curve). The obtained results are mainly the pressure deflexion recordings, strain measurements and visual observations of the geometrical changes. For thin heads, buckling is a very fast event and the first folding occurs sudently, with a strong perturbation on the pressure-deflexion curve. For the thickest heads, circular waves are slowly forming. In all of these tests, yielding occured before buckling and it was possible to increase the pressure beyond the first buckling pressure without failure. The experimental results agree very well (+-5% except one head) with the empirical formula Psub(c)=70000.(sigma y+sigma u/2)(t/D)sup(5/3)((D/H) 2 -8)sup(-2/3). The following notations being used: Psub(c): critical buckling pressure; sigma y: yield strength; sigma u: ultimate stress (same unit); t: knuckle thickness; D: mean diameter; H: height (same unit) [fr

  4. TESTING THE POSSIBLE INTRINSIC ORIGIN OF THE EXCESS VERY STRONG Mg II ABSORBERS ALONG GAMMA-RAY BURST LINES-OF-SIGHT

    International Nuclear Information System (INIS)

    Cucchiara, A.; Jones, T.; Charlton, J. C.; Fox, D. B.; Einsig, D.; Narayanan, A.

    2009-01-01

    The startling discovery by Prochter et al. that the frequency of very strong (W r (2796)>1 A) Mg II absorbers along gamma-ray burst (GRB) lines of sight ([dN/dz] GRB = 0.90) is more than three times the frequency along quasar lines of sight ([dN/dz] QSO = 0.24), over similar redshift ranges, has yet to be understood. In particular, explanations appealing to dust antibias in quasar samples, partial covering of the quasar sources, and gravitational-lensing amplification of the GRBs have all been carefully examined and found wanting. We therefore reconsider the possibility that the excess of very strong Mg II absorbers toward GRBs is intrinsic either to the GRBs themselves or to their immediate environment, and associated with bulk outflows with velocities as large as v max ∼ 0.3c. In order to examine this hypothesis, we accumulate a sample of 27 W r (2796)>1 A absorption systems found toward 81 quasars, and compare their properties to those of 8 W r (2796) > 1 A absorption systems found toward six GRBs; all systems have been observed at high spectral resolution (R = 45, 000) using the Ultraviolet and Visual Echelle Spectrograph on the Very Large Telescope. We make multiple comparisons of the absorber properties across the two populations, testing for differences in metallicity, ionization state, abundance patterns, dust abundance, kinematics, and phase structure. We find no significant differences between the two absorber populations using any of these metrics, implying that, if the excess of absorbers along GRB lines of sight are indeed intrinsic, they must be produced by a process which has strong similarities to the processes yielding strong Mg II systems associated with intervening galaxies. Although this may seem a priori unlikely, given the high outflow velocities required for any intrinsic model, we note that the same conclusion was reached, recently, with respect to the narrow absorption line systems seen in some quasars.

  5. Experimental results of direct containment heating by high-pressure melt ejection into the Surtsey vessel: The DCH-3 and DCH-4 tests

    International Nuclear Information System (INIS)

    Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W.; Nichols, R.T.; Sweet, D.W.

    1991-08-01

    Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a high-pressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris and oxygen. Thus, a comparison of the pressure response in DCH-3 and DCH-4 gave an indication of the DCH contribution due to metal/oxygen reactions. 44 refs., 110 figs., 43 tabs

  6. Experimental results of direct containment heating by high-pressure melt ejection into the Surtsey vessel: The DCH-3 and DCH-4 tests

    Energy Technology Data Exchange (ETDEWEB)

    Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W. (Sandia National Labs., Albuquerque, NM (United States)); Nichols, R.T. (Ktech Corp., Albuquerque, NM (United States)); Sweet, D.W. (AEA Technology, Winfrith (United Kingdom))

    1991-08-01

    Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a high-pressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris and oxygen. Thus, a comparison of the pressure response in DCH-3 and DCH-4 gave an indication of the DCH contribution due to metal/oxygen reactions. 44 refs., 110 figs., 43 tabs.

  7. Simulating X-ray bursts during a transient accretion event

    Science.gov (United States)

    Johnston, Zac; Heger, Alexander; Galloway, Duncan K.

    2018-06-01

    Modelling of thermonuclear X-ray bursts on accreting neutron stars has to date focused on stable accretion rates. However, bursts are also observed during episodes of transient accretion. During such events, the accretion rate can evolve significantly between bursts, and this regime provides a unique test for burst models. The accretion-powered millisecond pulsar SAX J1808.4-3658 exhibits accretion outbursts every 2-3 yr. During the well-sampled month-long outburst of 2002 October, four helium-rich X-ray bursts were observed. Using this event as a test case, we present the first multizone simulations of X-ray bursts under a time-dependent accretion rate. We investigate the effect of using a time-dependent accretion rate in comparison to constant, averaged rates. Initial results suggest that using a constant, average accretion rate between bursts may underestimate the recurrence time when the accretion rate is decreasing, and overestimate it when the accretion rate is increasing. Our model, with an accreted hydrogen fraction of X = 0.44 and a CNO metallicity of ZCNO = 0.02, reproduces the observed burst arrival times and fluences with root mean square (rms) errors of 2.8 h, and 0.11× 10^{-6} erg cm^{-2}, respectively. Our results support previous modelling that predicted two unobserved bursts and indicate that additional bursts were also missed by observations.

  8. High flux testing reactor Petten. Replacement of the reactor vessel and connected components. Overall report

    International Nuclear Information System (INIS)

    Chrysochoides, N.G.; Cundy, M.R.; Von der Hardt, P.; Husmann, K.; Swanenburg de Veye, R.J.; Tas, A.

    1985-01-01

    The project of replacing the HFR originated in 1974 when results of several research programmes confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report contains the detailed chronology of events concerning preparation and execution of the replacement. After a 14 months' outage the reactor resumed routine operation on 14th February, 1985. At the end of several years of planning and preparation the reconstruction proceded in the following steps: unloading of the old core, decay of short-lived radioactivity in December 1983, removal of the old tank and of its peripheral equipment in January-February 1984, segmentation and waste disposal of the removed components in March-April, decontamination of the pools, bottom penetration overhauling in May-June, installation of the new tank and other new components in July-September, testing and commissioning, including minor modifications in October-December, and, trials runs and start-up preparation in January-February 1985. The new HFR Petten features increased and improved experimental facilities. Among others the obsolete thermal columns was replaced by two high flux beam tubes. Moreover the new plant has been designed for future increases of reactor power and neutron fluxes. For the next three to four years the reactor has to cope with a large irradiation programme, claiming its capacity to nearly 100%

  9. ISP-50 Specifications for a Direct Vessel Injection Line Break Test with the ATLAS

    International Nuclear Information System (INIS)

    Choi, Ki Yong; Baek, Won Pil; Kim, Yeon Sik; Park, Hyun Sik; Cho, Seok; Kang, Kyoung Ho; Choi, Nam Hyun; Min, Kyoung Ho

    2009-06-01

    An OECD/NEA International Standard Problem Exercise (ISP) focussing on a DVI line break simulation result with the ATLAS was approved by the NEA Committee on the Safety of Nuclear Installation (CSNI) meeting in December 2008 and was numbered by ISP-50. The ISP-50 program will be operated by an operating agency, KAERI for three years starting from the physical year 2009. Fourteen international organizations confirmed their participation in the ISP-50, including NRC (USA), JAEA, JNES (Japan), GRS (Germany), KFKI-AEKI (Hungary), EDO Gidropress (Russia), VTT, Fortum (Finland), NRI (Czech Republic), Univ. of Pisa (Italy), KINS, KNF, KOPEC, and KAERI (Korea). In addition, KTH in Sweden and HSE in UK are considering late participation. Recently, NPIC and CIAE in China hope to join the ISP-50. As for the safety analysis codes, nine codes are expected to be used for the ISP-50: MARS-3D, RELAP5- 3D, RELAP5, TRACE, CATHARE, APROS, ATHELET, TRAP, and KORSAR. It is the first ISP exercise in Korea in which a domestic test facility is utilized by international nuclear society and this exercise will contribute to extending our physical understanding on thermal hydraulic phenomena during the DVI line break accidents and to verifying the best-estimate thermal-hydraulic safety analysis codes. This report was prepared to define technical specifications of the ISP-50 exercise according the guideline provided by OECD/CSNI. It includes general objectives, phases, deliverables to participants, parameters required for comparison and the time table

  10. Rapid sealing and cutting of porcine blood vessels, ex vivo, using a high-power, 1470-nm diode laser.

    Science.gov (United States)

    Giglio, Nicholas C; Hutchens, Thomas C; Perkins, William C; Latimer, Cassandra; Ward, Arlen; Nau, William H; Fried, Nathaniel M

    2014-03-01

    Suture ligation with subsequent cutting of blood vessels to maintain hemostasis during surgery is time consuming and skill intensive. Energy-based electrosurgical and ultrasonic devices are often used to replace sutures and mechanical clips to provide rapid hemostasis and decrease surgery time. Some of these devices may create undesirably large collateral zones of thermal damage and tissue necrosis, or require separate mechanical blades for cutting. Infrared lasers are currently being explored as alternative energy sources for vessel sealing applications. In a previous study, a 1470-nm laser was used to seal vessels 1 to 6 mm in diameter in 5 s, yielding burst pressures of ∼500  mmHg. The purpose of this study was to provide vessel sealing times comparable with current energy-based devices, incorporate transection of sealed vessels, and demonstrate high vessel burst pressures to provide a safety margin for future clinical use. A 110-W, 1470-nm laser beam was transmitted through a fiber and beam shaping optics, producing a 90-W linear beam 3.0 by 9.5 mm for sealing (400  W/cm2), and 1.1 by 9.6 mm for cutting (1080  W/cm2). A two-step process sealed and then transected ex vivo porcine renal vessels (1.5 to 8.5 mm diameter) in a bench top setup. Seal and cut times were 1.0 s each. A burst pressure system measured seal strength, and histologic measurements of lateral thermal spread were also recorded. All blood vessels tested (n=55 seal samples) were sealed and cut, with total irradiation times of 2.0 s and mean burst pressures of 1305±783  mmHg. Additional unburst vessels were processed for histological analysis, showing a lateral thermal spread of 0.94±0.48  mm (n=14 seal samples). This study demonstrated that an optical-based system is capable of precisely sealing and cutting a wide range of porcine renal vessel sizes and, with further development, may provide an alternative to radiofrequency- and ultrasonic-based vessel sealing devices.

  11. Use of Reactor Pressure Vessel Surveillance Materials for Extended Life Evaluations Using Power and Test Reactor Irradiations

    International Nuclear Information System (INIS)

    Server, W.L.; Nanstad, R.K.; Odette, G.R.

    2012-01-01

    The most important component in assuring safety of the nuclear power plant is the reactor pressure (RPV). Surveillance programs have been designed to cover the licensed life of operating nuclear RPVs. The original surveillance programs were designed when the licensed life was 40 years. More than one-half of the operating nuclear plants in the USA have an extended license out to 60 years, and there are plans to continue to operate many plants out to 80 years. Therefore, the surveillance programs have had to be adjusted or enhanced to generate key data for 60 years, and now consideration must be given for 80 or more years. To generate the necessary data to assure safe operation out to these extended license lives, test reactor irradiations have been initiated with key RPV and model alloy steels, which include several steels irradiated in the current power reactor surveillance programs out to relatively high fluence levels. These data are crucial in understanding the radiation embrittlement mechanisms and to enable extrapolation of the irradiation effects on mechanical properties for these extended time periods. This paper describes the potential radiation embrittlement mechanisms and effects when assessing much longer operating times and higher neutron fluence levels. Potential methods for adjusting higher neutron flux test reactor data for use in predicting power reactor vessel conditions are discussed. (author)

  12. Testing a polarimetric cloud imager aboard research vessel Polarstern: comparison of color-based and polarimetric cloud detection algorithms.

    Science.gov (United States)

    Barta, András; Horváth, Gábor; Horváth, Ákos; Egri, Ádám; Blahó, Miklós; Barta, Pál; Bumke, Karl; Macke, Andreas

    2015-02-10

    Cloud cover estimation is an important part of routine meteorological observations. Cloudiness measurements are used in climate model evaluation, nowcasting solar radiation, parameterizing the fluctuations of sea surface insolation, and building energy transfer models of the atmosphere. Currently, the most widespread ground-based method to measure cloudiness is based on analyzing the unpolarized intensity and color distribution of the sky obtained by digital cameras. As a new approach, we propose that cloud detection can be aided by the additional use of skylight polarization measured by 180° field-of-view imaging polarimetry. In the fall of 2010, we tested such a novel polarimetric cloud detector aboard the research vessel Polarstern during expedition ANT-XXVII/1. One of our goals was to test the durability of the measurement hardware under the extreme conditions of a trans-Atlantic cruise. Here, we describe the instrument and compare the results of several different cloud detection algorithms, some conventional and some newly developed. We also discuss the weaknesses of our design and its possible improvements. The comparison with cloud detection algorithms developed for traditional nonpolarimetric full-sky imagers allowed us to evaluate the added value of polarimetric quantities. We found that (1) neural-network-based algorithms perform the best among the investigated schemes and (2) global information (the mean and variance of intensity), nonoptical information (e.g., sun-view geometry), and polarimetric information (e.g., the degree of polarization) improve the accuracy of cloud detection, albeit slightly.

  13. PWR clad ballooning: The effect of circumferential clad temperature variations on the burst strain/burst temperature relationship

    International Nuclear Information System (INIS)

    Barlow, P.

    1983-01-01

    By experiment, it has been shown by other workers that there is a reduction in the creep ductility of Zircaloy 4 in the α+β phase transition region. Results from single rod burst tests also show a reduction in burst strain in the α+β phase region. In this report it is shown theoretically that for single rod burst tests in the presence of circumferential temperature gradients, the temperature dependence of the mean burst strain is not determined by temperature variations in creep ductility, but is governed by the temperature sensitivity of the creep strain rate, which is shown to be a maximum in the α+β phase transition region. To demonstrate this effect, the mean clad strain at burst was calculated for creep straining at different temperature levels in the α, α+β and β phase regions. Cross-pin temperature gradients were applied which produced strain variations around the clad which were greatest in the α+β phase region. The mean strain at burst was determined using a maximum local burst strain (i.e. a creep ductility) which is independent of temperature. By assuming cross-pin temperature gradients which are typical of those observed during burst tests, then the calculated mean burst strain/burst temperature relationship gave good agreement with experiment. The calculations also show that when circumferential temperature differences are present, the calculated mean strain at burst is not sensitive to variations in the magnitude of the assumed creep ductility. This reduces the importance of the assumed burst criterion in the calculations. Hence a temperature independent creep ductility (e.g. 100% local strain) is adequate as a burst criterion for calculations under PWR LOCA conditions. (author)

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

    International Nuclear Information System (INIS)

    Domian, H.A.

    1987-02-01

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

  15. On results of tests of thermal insulation structural fragments for in-vessel equipment and pipelines of the VG-400 plant on vibrational and acoustic loads

    International Nuclear Information System (INIS)

    Ledenko, S.A.; Andreev, V.A.; Mirenkov, A.F.; Zakharov, V.A.; Suvorov, V.E.; Prokimnov, V.V.

    1990-01-01

    Results of vibrostrength and acoustic fatigue tests of the fragments of thermal insulation for in-vessel equipment and pipelines of the VG-400 reactor are presented. The insulation structure is based on the insulation layer made of steel foil and carbon materials. Weak points in the insulation structure, namely - the welded joints of stiffening ribs - are detected in the course of testing. A conclusion is made on the possibility of vibrational test substitution for the acoustic ones

  16. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (Capsule 2) of Ulchin Nuclear Power Plant Unit 4

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2007-04-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin Unit 4 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.306E+18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.918 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.615E+18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 8.478E+18 and 1.673E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin Unit 4 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  17. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (CAPSULE 2) of Ulchin Nuclear Power Plant Unit 3

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2006-12-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.674E 18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.920 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.913E 18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 9.249E 18 and 1.834E 19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  18. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, Masataka; Takeda, Nobukazu; Onozuka, Masanori [Japan Atomic Energy Agency (Japan); Kakudate, Satoshi [Mitsubishi Heavy Industries, Ltd. (Japan)

    2007-07-01

    The structure and fabrication methods of the ITER vacuum vessel have been investigated and defined by the ITER international team. However, some of the current specifications are very difficult to be achieved from the manufacturing point of view and will lead to cost increase. In the mock-up fabrication, it is planned to conduct the following items: 1. Feasibility of the Japanese proposed VV structure and fabrication methods and the applicability to the ITER are to be confirmed; 2. Assembly procedure and inspection procedure are to be confirmed; 3. Manufacturing tolerances are to be assessed; 4. Manufacturing schedule is to be assessed. This report summarizes the Japanese proposed specification of the VV mock-up describing differences between the ITER supplied design. General scope of the mock-up fabrication and the detailed dimensions are also shown. In the VV fabrication, several types of weld joint configuration will be used. This report shows the joint configurations proposed by Japan to be used for the inner shell connection, the rib-to-shell connection and outer shell connection, and the housing-to-shell connection, respectively. Non-destructive testing considered to be applied to each joint configuration is also presented. A series of the fabrication and assembly procedures for the mock-up are presented in this report, together with candidates of welding configurations. Finally, the report summarizes the results of mock-up fabrication, including results of nondestructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. (orig.)

  19. The Drift Burst Hypothesis

    DEFF Research Database (Denmark)

    Christensen, Kim; Oomen, Roel; Renò, Roberto

    are an expected and regular occurrence in financial markets that can arise through established mechanisms such as feedback trading. At a theoretical level, we show how to build drift bursts into the continuous-time Itô semi-martingale model in such a way that the fundamental arbitrage-free property is preserved......, currencies and commodities. We find that the majority of identified drift bursts are accompanied by strong price reversals and these can therefore be regarded as “flash crashes” that span brief periods of severe market disruption without any material longer term price impacts....

  20. Pressurized Vessel Slurry Pumping

    International Nuclear Information System (INIS)

    Pound, C.R.

    2001-01-01

    This report summarizes testing of an alternate ''pressurized vessel slurry pumping'' apparatus. The principle is similar to rural domestic water systems and ''acid eggs'' used in chemical laboratories in that material is extruded by displacement with compressed air

  1. Cosmic gamma-ray burst

    International Nuclear Information System (INIS)

    Yamagami, Takamasa

    1985-01-01

    Ballon experiments for searching gamma-ray burst were carried out by employing rotating-cross modulation collimators. From a very long observation of total 315 hours during 1975 to 1979, three gamma-ray intensity anomalies were observed which were speculated as a gamma-ray burst. As for the first gamma-ray intensity anomaly observed in 1975, the burst source could be located precisely but the source, heavenly body, could not be specified. Gamma-ray burst source estimation was made by analyzing distribution of burst source in the celestial sphere, burst size distribution, and burst peak. Using the above-mentioned data together with previously published ones, apparent inconsistency was found between the observed results and the adopted theory that the source was in the Galaxy, and this inconsistency was found due to the different time profiles of the burst observed with instruments of different efficiency. It was concluded by these analysis results that employment of logN - logP (relation between burst frequency and burst count) was better than that of logN - logS (burst size) in the examination of gamma-ray burst because the former was less uncertain than the latter. Analyzing the author's observed gamma-ray burst data and the related published data, it was clarified that the burst distribution was almost P -312 for the burst peak value larger than 10 -6 erg/cm 2 .sec. The author could indicate that the calculated celestial distribution of burst source was consistent with the observed results by the derivation using the logN - logP relationship and that the burst larger than 10 -6 erg/cm 2 .sec happens about one thousand times a year, about ten times of the previous value. (Takagi, S.)

  2. Development of advanced manufacturing technologies for low cost hydrogen storage vessels

    Energy Technology Data Exchange (ETDEWEB)

    Leavitt, Mark [Quantum Fuel Systems Technologies Worldwide, Inc., Irvine, CA (United States); Lam, Patrick [Boeing Research and Technology (BR& T), Seattle, WA (United States)

    2014-12-29

    The U.S. Department of Energy (DOE) defined a need for low-cost gaseous hydrogen storage vessels at 700 bar to support cost goals aimed at 500,000 units per year. Existing filament winding processes produce a pressure vessel that is structurally inefficient, requiring more carbon fiber for manufacturing reasons, than would otherwise be necessary. Carbon fiber is the greatest cost driver in building a hydrogen pressure vessel. The objective of this project is to develop new methods for manufacturing Type IV pressure vessels for hydrogen storage with the purpose of lowering the overall product cost through an innovative hybrid process of optimizing composite usage by combining traditional filament winding (FW) and advanced fiber placement (AFP) techniques. A numbers of vessels were manufactured in this project. The latest vessel design passed all the critical tests on the hybrid design per European Commission (EC) 79-2009 standard except the extreme temperature cycle test. The tests passed include burst test, cycle test, accelerated stress rupture test and drop test. It was discovered the location where AFP and FW overlap for load transfer could be weakened during hydraulic cycling at 85°C. To design a vessel that passed these tests, the in-house modeling software was updated to add capability to start and stop fiber layers to simulate the AFP process. The original in-house software was developed for filament winding only. Alternative fiber was also investigated in this project, but the added mass impacted the vessel cost negatively due to the lower performance from the alternative fiber. Overall the project was a success to show the hybrid design is a viable solution to reduce fiber usage, thus driving down the cost of fuel storage vessels. Based on DOE’s baseline vessel size of 147.3L and 91kg, the 129L vessel (scaled to DOE baseline) in this project shows a 32% composite savings and 20% cost savings when comparing Vessel 15 hybrid design and the Quantum

  3. Nanolensed Fast Radio Bursts

    Science.gov (United States)

    Eichler, David

    2017-12-01

    It is suggested that fast radio bursts can probe gravitational lensing by clumpy dark matter objects that range in mass from 10-3 M ⊙-102 M ⊙. They may provide a more sensitive probe than observations of lensings of objects in the Magellanic Clouds, and could find or rule out clumpy dark matter with an extended mass spectrum.

  4. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    International Nuclear Information System (INIS)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young

    2016-01-01

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation

  5. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation.

  6. Cyclic Crack Growth Testing of an A.O. Smith Multilayer Pressure Vessel with Modal Acoustic Emission Monitoring and Data Assessment

    Science.gov (United States)

    Ziola, Steven M.

    2014-01-01

    Digital Wave Corp. (DWC) was retained by Jacobs ATOM at NASA Ames Research Center to perform cyclic pressure crack growth sensitivity testing on a multilayer pressure vessel instrumented with DWC's Modal Acoustic Emission (MAE) system, with captured wave analysis to be performed using DWCs WaveExplorerTM software, which has been used at Ames since 2001. The objectives were to document the ability to detect and characterize a known growing crack in such a vessel using only MAE, to establish the sensitivity of the equipment vs. crack size and / or relevance in a realistic field environment, and to obtain fracture toughness materials properties in follow up testing to enable accurate crack growth analysis. This report contains the results of the testing.

  7. Application of acoustic emission monitoring to pressure tests of a steam receiver vessel with flawed nozzle welds

    International Nuclear Information System (INIS)

    Woodward, B.; McDonald, N.R.; Hincksman, M.J.

    1976-01-01

    As part of the first stage of an Australian Welding Research Association co-operative research project, acoustic emission monitoring has been applied to a steam receiver vessel withdrawn from service owing to severe weld cracking. This technique is used to check acceptance standards for defects in nozzle welds and to apply modern methods of assessing the integrity of pressurised plant. Acoustic emission monitoring has been used, together with strain gauge measurements and ultrasonic scanning, to detect the occurrence of any significant defect growth during cyclic pressurisation of the vessel. During this first stage, no significant defect growth has been produced by 1000 cycles of pressure up to 24.1 MPa (3500 psi), subsequent pressurisation up to 35.8 MPa (5200 psi), or 97 per cent of the expected yield stress of the vessel shell. The small amount of acoustic emission detected was consistent with this result. (author)

  8. Tempest in a vessel

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-01-01

    As the ASN made some statements about anomalies of carbon content in the EPR vessel bottom and top, the author recalls and comments some technical issues to better understand the information published on this topic. He notably addresses the role of the vessel, briefly indicates its operating conditions, shape and structure, and mechanical components for the top, its material and mechanical properties, and test samples used to assess mechanical properties. He also comments the phenomenon of radio-induced embrittlement, the vessel manufacturing process, and evokes the applicable regulations. He quotes and comments statements made by the ASN and Areva which evoke further assessments of the concerned components

  9. Scientific Applications Performance Evaluation on Burst Buffer

    KAUST Repository

    Markomanolis, George S.

    2017-10-19

    Parallel I/O is an integral component of modern high performance computing, especially in storing and processing very large datasets, such as the case of seismic imaging, CFD, combustion and weather modeling. The storage hierarchy includes nowadays additional layers, the latest being the usage of SSD-based storage as a Burst Buffer for I/O acceleration. We present an in-depth analysis on how to use Burst Buffer for specific cases and how the internal MPI I/O aggregators operate according to the options that the user provides during his job submission. We analyze the performance of a range of I/O intensive scientific applications, at various scales on a large installation of Lustre parallel file system compared to an SSD-based Burst Buffer. Our results show a performance improvement over Lustre when using Burst Buffer. Moreover, we show results from a data hierarchy library which indicate that the standard I/O approaches are not enough to get the expected performance from this technology. The performance gain on the total execution time of the studied applications is between 1.16 and 3 times compared to Lustre. One of the test cases achieved an impressive I/O throughput of 900 GB/s on Burst Buffer.

  10. Concept of a Prestressed Cast Iron Pressure Vessel for a Modular High Temperature Reactor

    International Nuclear Information System (INIS)

    Steinwarz, Wolfgang; Bounin, Dieter

    2014-01-01

    High Temperature Reactors (HTR) are representing one of the most interesting solutions for the upcoming generation of nuclear technology, especially with view to their inherent safety characteristics. To complete the safety concept of such plants already in the first phase of the technical development, Prestressed Cast Iron Pressure Vessels (PCIV) instead of the established forged steel reactor pressure vessels have been considered under the aspect of safety against bursting. A longterm research and development work, mainly performed in Germany, showed the excellent features of this technical solution. Diverse prototypic vessels were tested and officially proven. Design studies confirmed the feasibility of such a vessel concept also for Light Water Reactor types, too. The main concept elements of such a burst-proof vessel are: Strength and tightness functions are structurally separated. The tensile forces are carried by the prestressing systems consisting of a large number of independent wires. Compressive forces are applied to the vessel walls and heads. These are segmented into blocks of ductile cast iron. All cast iron blocks are prestressed to high levels of compression. The sealing function is assigned to a steel liner fixed to the cast iron blocks. The prestressing system is designed for an ultimate pressure of 2.3 times the design pressure. The prestress of the lids is designed for gapping at a much smaller pressure. Therefore, a drop of pressure will always occur before loss of strength (“leakage before failure”). In addition to these safety features further technical as well as economic aspects generate favorable assessment criteria: high design flexibility, feasibility of large vessel diameters; advantageous conditions for transport, assembly and decommissioning due to the segmented construction; advantage of workshop manufacturing; high-level quality control of components. Nowadays, considering the globally newly standardized safety requirements

  11. Fast Radio Burst/Gamma-Ray Burst Cosmography

    Science.gov (United States)

    Gao, He; Li, Zhuo; Zhang, Bing

    2014-06-01

    Recently, both theoretical arguments and observational evidence suggested that a small fraction of fast radio bursts (FRBs) could be associated with gamma-ray bursts (GRBs). If such FRB/GRB association systems are commonly detected in the future, the combination of dispersion measures (DM) derived from FRBs and redshifts derived from GRBs makes these systems a plausible tool to conduct cosmography. We quantify uncertainties in deriving the redshift-dependent DM_{IGM} as a function of z and test how well dark energy models can be constrained with Monte Carlo simulations. We show that with several tens of FRB/GRB systems potentially detected in a decade or so, one may reach reasonable constraints on wCDM models. When combined with Type Ia supernova (SN Ia) data, unprecedented constraints on the dark energy equation of state may be achieved, thanks to the prospects of detecting FRB/GRB systems at relatively high redshifts. The ratio between the mean value \\lt {DM_IGM} (z)\\gt and luminosity distance (D L(z)) is insensitive to dark energy models. This gives the prospect of applying SN Ia data to calibrate \\lt {DM_IGM} (z)\\gt using a relatively small sample of FRB/GRB systems, allowing a reliable constraint on the baryon inhomogeneity distribution as a function of redshift. The methodology developed in this paper can also be applied if the FRB redshifts can be measured by other means. Some caveats of putting this method into practice are also discussed.

  12. Fast radio burst/gamma-ray burst cosmography

    International Nuclear Information System (INIS)

    Gao, He; Zhang, Bing; Li, Zhuo

    2014-01-01

    Recently, both theoretical arguments and observational evidence suggested that a small fraction of fast radio bursts (FRBs) could be associated with gamma-ray bursts (GRBs). If such FRB/GRB association systems are commonly detected in the future, the combination of dispersion measures (DM) derived from FRBs and redshifts derived from GRBs makes these systems a plausible tool to conduct cosmography. We quantify uncertainties in deriving the redshift-dependent DM IGM as a function of z and test how well dark energy models can be constrained with Monte Carlo simulations. We show that with several tens of FRB/GRB systems potentially detected in a decade or so, one may reach reasonable constraints on wCDM models. When combined with Type Ia supernova (SN Ia) data, unprecedented constraints on the dark energy equation of state may be achieved, thanks to the prospects of detecting FRB/GRB systems at relatively high redshifts. The ratio between the mean value and luminosity distance (D L (z)) is insensitive to dark energy models. This gives the prospect of applying SN Ia data to calibrate using a relatively small sample of FRB/GRB systems, allowing a reliable constraint on the baryon inhomogeneity distribution as a function of redshift. The methodology developed in this paper can also be applied if the FRB redshifts can be measured by other means. Some caveats of putting this method into practice are also discussed.

  13. Fast radio burst/gamma-ray burst cosmography

    Energy Technology Data Exchange (ETDEWEB)

    Gao, He; Zhang, Bing [Department of Physics and Astronomy, University of Nevada Las Vegas, NV 89154 (United States); Li, Zhuo, E-mail: gaohe@physics.unlv.edu, E-mail: zhang@physics.unlv.edu, E-mail: zhuo.li@pku.edu.cn [Department of Astronomy, School of Physics, Peking University, Beijing 100871 (China)

    2014-06-20

    Recently, both theoretical arguments and observational evidence suggested that a small fraction of fast radio bursts (FRBs) could be associated with gamma-ray bursts (GRBs). If such FRB/GRB association systems are commonly detected in the future, the combination of dispersion measures (DM) derived from FRBs and redshifts derived from GRBs makes these systems a plausible tool to conduct cosmography. We quantify uncertainties in deriving the redshift-dependent DM{sub IGM} as a function of z and test how well dark energy models can be constrained with Monte Carlo simulations. We show that with several tens of FRB/GRB systems potentially detected in a decade or so, one may reach reasonable constraints on wCDM models. When combined with Type Ia supernova (SN Ia) data, unprecedented constraints on the dark energy equation of state may be achieved, thanks to the prospects of detecting FRB/GRB systems at relatively high redshifts. The ratio between the mean value and luminosity distance (D {sub L}(z)) is insensitive to dark energy models. This gives the prospect of applying SN Ia data to calibrate using a relatively small sample of FRB/GRB systems, allowing a reliable constraint on the baryon inhomogeneity distribution as a function of redshift. The methodology developed in this paper can also be applied if the FRB redshifts can be measured by other means. Some caveats of putting this method into practice are also discussed.

  14. The development of a burst criterion for zircaloy fuel cladding under LOCA conditions

    International Nuclear Information System (INIS)

    Neitzel, H.J.; Rossinger, H.E.

    1980-02-01

    A burst criterion model, which assumes that deformation is controlled by steady-state creep, has been developed for a thin-walled cladding, in this case Zircaloy-4, subjected to a differential pressure and high temperature. The creep equation is integrated to obtain a burst time at the singularity of the strain. Once the burst time is known, the burst temperature and burst pressure can be calculated from the known temperature and pressure histories. A further relationship between burst stress and burst temperature is used to calculate the burst strain. Comparison with measured burst data shows good agreement between theory and experiment was found that, if the heating rate is constant, the burst temperature increases with decreasing stress, and that, if the stress level is constant, the burst temperature increases with increasing heating rate. It was also found that anisotropy alters the burst temperature and burst strain, and that test conditions in the α-Zr temperature range have no influence on the burst data. (auth)

  15. A repeating fast radio burst.

    Science.gov (United States)

    Spitler, L G; Scholz, P; Hessels, J W T; Bogdanov, S; Brazier, A; Camilo, F; Chatterjee, S; Cordes, J M; Crawford, F; Deneva, J; Ferdman, R D; Freire, P C C; Kaspi, V M; Lazarus, P; Lynch, R; Madsen, E C; McLaughlin, M A; Patel, C; Ransom, S M; Seymour, A; Stairs, I H; Stappers, B W; van Leeuwen, J; Zhu, W W

    2016-03-10

    Fast radio bursts are millisecond-duration astronomical radio pulses of unknown physical origin that appear to come from extragalactic distances. Previous follow-up observations have failed to find additional bursts at the same dispersion measure (that is, the integrated column density of free electrons between source and telescope) and sky position as the original detections. The apparent non-repeating nature of these bursts has led to the suggestion that they originate in cataclysmic events. Here we report observations of ten additional bursts from the direction of the fast radio burst FRB 121102. These bursts have dispersion measures and sky positions consistent with the original burst. This unambiguously identifies FRB 121102 as repeating and demonstrates that its source survives the energetic events that cause the bursts. Additionally, the bursts from FRB 121102 show a wide range of spectral shapes that appear to be predominantly intrinsic to the source and which vary on timescales of minutes or less. Although there may be multiple physical origins for the population of fast radio bursts, these repeat bursts with high dispersion measure and variable spectra specifically seen from the direction of FRB 121102 support an origin in a young, highly magnetized, extragalactic neutron star.

  16. Development of Mini-Compact Tension Test Method for Determining Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-05-01

    Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Validation of the mini compact tension specimen (mini-CT) geometry has been performed on previously well characterized Midland beltline Linde 80 (WF-70) weld in the unirradiated condition. It was shown that the fracture toughness transition temperature, To, measured by these Mini-CT specimens is almost the same as To value that was derived from various larger fracture toughness specimens. Moreover, an International collaborative program has been established to extend the assessment and validation efforts to irradiated Linde 80 weld metal. The program is underway and involves the Oak Ridge National Laboratory (ORNL), Central Research Institute for Electrical Power Industry (CRIEPI), and Electric Power Research Institute (EPRI). The irradiated Mini-CT specimens from broken halves of previously tested Charpy

  17. Gamma-ray bursts

    CERN Document Server

    Wijers, Ralph A M J; Woosley, Stan

    2012-01-01

    Cosmic gamma ray bursts (GRBs) have fascinated scientists and the public alike since their discovery in the late 1960s. Their story is told here by some of the scientists who participated in their discovery and, after many decades of false starts, solved the problem of their origin. Fourteen chapters by active researchers in the field present a detailed history of the discovery, a comprehensive theoretical description of GRB central engine and emission models, a discussion of GRB host galaxies and a guide to how GRBs can be used as cosmological tools. Observations are grouped into three sets from the satellites CGRO, BeppoSAX and Swift, and followed by a discussion of multi-wavelength observations. This is the first edited volume on GRB astrophysics that presents a fully comprehensive review of the subject. Utilizing the latest research, Gamma-ray Bursts is an essential desktop companion for graduate students and researchers in astrophysics.

  18. Tests of the geometrical description of blood vessels in a thermal model using counter-current geometries

    NARCIS (Netherlands)

    van Leeuwen, G. M.; Kotte, A. N.; Crezee, J.; Lagendijk, J. J.

    1997-01-01

    We have developed a thermal model, for use in hyperthermia treatment planning, in which blood vessels are described as geometrical objects; 3D curves with associated diameters. For the calculation of the heat exchange with the tissue an analytic result is used. To arrive at this result some

  19. Final report for the 5th surveillance test of the reactor pressure vessel material (capsule Y) of Yonggwang Nuclear Power Plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sam Lai; Kim, ByoungChul; Chang, Kee Ok (and others)

    2006-02-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X, W and Y are 5.777E+18, 1.5371E+19, 3.7634E+19, 4.3045E+19, and 4.8662E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.953 for the 1st through 5th testing and the calculational uncertainty,7.2% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.659E+19n/cm{sup 2} based on the end of 13th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 48, 56 and 64EFPY would reach 3.625E+19, 5.293E+19, 6.127E+19 and 6.960E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the Pressurized Thermal Shock(PTS) during the operation until design life.

  20. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Younggwang nuclear power plant unit1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 1 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.555E+18, 1.662E+19, 3.358E+19, and 4.521E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.859 for the 1st through 4th testing and the calculational uncertainty, 11.80% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.551E+19 n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.929E+19, 4.880E+19, 5.831E+19 and 6.782E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 1 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 4 refs., 41 figs., 35 tabs. (Author)

  1. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Yonggwang nuclear power plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-02-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.762E+18, 1.5391E+19, 3.5119E+19, and 4.2610E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.899 for the 1st through 4th testing and the calculational uncertainty, 12.3% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.357E+19 n/cm{sup 2} based on the end of 11th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.525E+19, 4.337E+19, 5.148E+19 and 5.960E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 42 tabs. (Author)

  2. Licensing experiences, risk assessment, demonstration test on nuclear fuel packages and design criteria for sea going vessel carrying spent fuel in Japan

    International Nuclear Information System (INIS)

    Aoki, S.; Ikeda, K.

    1978-01-01

    In Japan spent fuels from nuclear power plants shall be shipped to reprocessing plants by sea-going vessels. Atomic Energy Committee has initiated a board of experts to implement the assessment of environmental safety for sea transport. As a part of the assessment a study has been conducted by Central Research Institute of Electric Power Industry under sponsorship of Nuclear Safety Bureau, which is intended to guarantee the safety of sea transport. Nuclear Safety Bureau also has a program to carry out a long term demonstration test on spent fuel package using full scale package models. The test consists of drop, heat transfer, fire, collapse under high external pressure, immersion, shielding and subcritical test. The purpose of this test is to obtain the public acceptance and also to verify the adequacy of the safety analysis for nuclear fuel packages. In order to secure the safety of sea transport, the Ministry of Transportation has provided for the design criteria for sea-going vessel in the case of full load shipping, which aims to make minimum the probability of sinking at collision, grounding and other unforeseen accidents on the sea and also to retain the radiation exposure to crews as low as possible. The design criteria consists of the following items: (1) structural strength of vessel, (2) collision protective structure, (3) arrangement of holds, (4) stability after damage, (5) grounding protective structure, (6) cooling system, (7) tie-down equipment, (8) radiation inspection apparatus, (9) decontamination facilities, (10) emergency water flooding equipment for ship fire, (11) emergency electric sources, etc. Based on the design criteria a sea-going vessel names HINOURA-MARU has been reconstructed to transport spent fuel packages from nuclear power stations to the reprocessing plant

  3. Simulation and tests to individual and coupled models of the reactor vessel simulator and the recirculation system for the SUN-RAH

    International Nuclear Information System (INIS)

    Sanchez S, R.A.

    2004-01-01

    The present project, is continuation of the project presented in the congress SNM-2003. In this new phase of the project, they were carried out adaptive changes to the modeling and implementation of the module of the full superior of the core of the reactor, they were carried out those modeling of the generation of heat as well as of the energy transfer in the one fuel. These models present the main characteristics of the vessel of the one reactor and of the recirculation system, defined by the main phenomena that they intervene in the physical processes, in the previous version the simulation in real time it required of an extremely quick computer and without executing collateral processes. The tests are presented carried out to the different models belonging to the Simulator of the Reactor Vessel and the Recirculation system for the SUN-RAH (University Simulator of Nucleo electric with Boiling Water Reactor), as well as the results hurtled by this tests. In each section the executions of the tests and the corresponding analyses of results are shown for each pattern. Besides the above mentioned, the advantages presented by the Simulator of the reactor vessel and the recirculation system are pointed. (Author)

  4. High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi

    1999-02-01

    In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10 6 R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

  5. High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan)] [and others

    1999-02-01

    In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10{sup 6} R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

  6. Empirical Constraints on the Origin of Fast Radio Bursts: Volumetric Rates and Host Galaxy Demographics as a Test of Millisecond Magnetar Connection

    Science.gov (United States)

    Nicholl, M.; Williams, P. K. G.; Berger, E.; Villar, V. A.; Alexander, K. D.; Eftekhari, T.; Metzger, B. D.

    2017-07-01

    The localization of the repeating fast radio burst (FRB) 121102 to a low-metallicity dwarf galaxy at z = 0.193, and its association with a luminous quiescent radio source, suggests the possibility that FRBs originate from magnetars, formed by the unusual supernovae that occur in such galaxies. We investigate this possibility via a comparison of magnetar birth rates, the FRB volumetric rate, and host galaxy demographics. We calculate average volumetric rates of possible millisecond magnetar production channels, such as superluminous supernovae (SLSNe), long and short gamma-ray bursts (GRBs), and general magnetar production via core-collapse supernovae (CCSNe). For each channel, we also explore the expected host galaxy demographics using their known properties. We determine for the first time the number density of FRB emitters (the product of their volumetric birth rate and lifetime), {R}{FRB}τ ≈ {10}4 Gpc-3, assuming that FRBs are predominantly emitted from repetitive sources similar to FRB 121102 and adopting a beaming factor of 0.1. By comparing rates, we find that production via rare channels (SLSNe, GRBs) implies a typical FRB lifetime of ˜30-300 years, in good agreement with other lines of argument. The total energy emitted over this time is consistent with the available energy stored in the magnetic field. On the other hand, any relation to magnetars produced via normal CCSNe leads to a very short lifetime of ˜0.5 years, in conflict with both theory and observation. We demonstrate that due to the diverse host galaxy distributions of the different progenitor channels, many possible sources of FRB birth can be ruled out with ≲ 10 host galaxy identifications. Conversely, targeted searches of galaxies that have previously hosted decades-old SLSNe and GRBs may be a fruitful strategy for discovering new FRBs and related quiescent radio sources, and determining the nature of their progenitors.

  7. The study of the irradiation-induced embrittlement of reactor pressure vessels. Analysis of surveillance test specimens of a commercial nuclear reactor pressure vessel studied by three-dimensional atom probe and positron annihilation

    International Nuclear Information System (INIS)

    Nagai, Yasuyoshi; Toyama, Takeshi; Hasegawa, Masayuki

    2007-01-01

    The study of embrittlement of nuclear power reactor pressure vessels (RPVs) is of critical importance for the safety assessment in the nuclear industry. Some origins of embrittlement are attributed to fine Cu precipitates, matrix defects, grain boundary segregation of P and late blooming phase. This review article described nanostructural observation by three-dimensional atom probe (3DAP) and positron annihilation spectroscopy (PAS). The density and sizes of Cu-rich nanoprecipitates and grain boundary segregation are sensitively detected by 3DAP, and vacancies are probed by PAS. Element analysis around vacancies and fine microstructural Cu precipitates not containing vacancies are successfully observed by a coincidence doppler broadening method. The nanostructural evolution of irradiation-induced Cu-rich nanoprecipitates (CRNPs) and vacancy clusters in surveillance test specimens of commercial nuclear reactor pressure vessel steel welds of Doel-2 in Belgium were revealed by combining 3DAP and PAS. In both medium (0.13 wt%) and high (0.30 wt%) Cu welds, the CRNPs were found to form readily at the very beginning of the reactor lifetime. On the other hand, small vacancy clusters start appearing after the initial Cu precipitates and accumulate steadily with increasing neutron dose. The CRNPs were also observed at very low dose rate of neutrons in the test specimen of Calder Hall Reactor of Japan Atomic Power Company. The significant enhancement of these Cu precipitates results in the embrittlement in practical RPVs. At very high dose of 2.2x10 18 n/cm 2 by JMTR, the Cu precipitates were scarcely observed, and the irradiation-induced embrittlement was primarily caused from vacancy-impurity complexes and dislocation loops. (author)

  8. Gamma-Ray Bursts

    Science.gov (United States)

    Pellizza, L. J.

    Gamma-ray bursts are the brightest transient sources in the gamma-ray sky. Since their discovery in the late 1960s, the investigation of the astrophysical sys- tems in which these phenomena take place, and the physical mechanisms that drive them, has become a vast and prolific area of modern astrophysics. In this work I will briefly describe the most relevant observations of these sources, and the models that describe their nature, emphasizing on the in- vestigations about the progenitor astrophysical systems. FULL TEXT IN SPANISH

  9. Gamma Ray Bursts

    Science.gov (United States)

    Gehrels, Neil; Meszaros, Peter

    2012-01-01

    Gamma-ray bursts (GRBs) are bright flashes of gamma-rays coming from the cosmos. They occur roughly once per day ,last typically lOs of seconds and are the most luminous events in the universe. More than three decades after their discovery, and after pioneering advances from space and ground experiments, they still remain mysterious. The launch of the Swift and Fermi satellites in 2004 and 2008 brought in a trove of qualitatively new data. In this review we survey the interplay between these recent observations and the theoretical models of the prompt GRB emission and the subsequent afterglows.

  10. Endogenous GABA and Glutamate Finely Tune the Bursting of Olfactory Bulb External Tufted Cells

    Science.gov (United States)

    Hayar, Abdallah; Ennis, Matthew

    2008-01-01

    In rat olfactory bulb slices, external tufted (ET) cells spontaneously generate spike bursts. Although ET cell bursting is intrinsically generated, its strength and precise timing may be regulated by synaptic input. We tested this hypothesis by analyzing whether the burst properties are modulated by activation of ionotropic γ-aminobutyric acid (GABA) and glutamate receptors. Blocking GABAA receptors increased—whereas blocking ionotropic glutamate receptors decreased—the number of spikes/burst without changing the interburst frequency. The GABAA agonist (isoguvacine, 10 μM) completely inhibited bursting or reduced the number of spikes/burst, suggesting a shunting effect. These findings indicate that the properties of ET cell spontaneous bursting are differentially controlled by GABAergic and glutamatergic fast synaptic transmission. We suggest that ET cell excitatory and inhibitory inputs may be encoded as a change in the pattern of spike bursting in ET cells, which together with mitral/tufted cells constitute the output circuit of the olfactory bulb. PMID:17567771

  11. Gamma-ray burst spectra

    International Nuclear Information System (INIS)

    Teegarden, B.J.

    1982-01-01

    A review of recent results in gamma-ray burst spectroscopy is given. Particular attention is paid to the recent discovery of emission and absorption features in the burst spectra. These lines represent the strongest evidence to date that gamma-ray bursts originate on or near neutron stars. Line parameters give information on the temperature, magnetic field and possibly the gravitational potential of the neutron star. The behavior of the continuum spectrum is also discussed. A remarkably good fit to nearly all bursts is obtained with a thermal-bremsstrahlung-like continuum. Significant evolution is observed of both the continuum and line features within most events

  12. UWB dual burst transmit driver

    Science.gov (United States)

    Dallum, Gregory E [Livermore, CA; Pratt, Garth C [Discovery Bay, CA; Haugen, Peter C [Livermore, CA; Zumstein, James M [Livermore, CA; Vigars, Mark L [Livermore, CA; Romero, Carlos E [Livermore, CA

    2012-04-17

    A dual burst transmitter for ultra-wideband (UWB) communication systems generates a pair of precisely spaced RF bursts from a single trigger event. An input trigger pulse produces two oscillator trigger pulses, an initial pulse and a delayed pulse, in a dual trigger generator. The two oscillator trigger pulses drive a gated RF burst (power output) oscillator. A bias driver circuit gates the RF output oscillator on and off and sets the RF burst packet width. The bias driver also level shifts the drive signal to the level that is required for the RF output device.

  13. Bipolar impedance-controlled sealing of the pulmonary artery with SealSafe G3 electric current: determination of bursting pressures in an ex vivo model.

    Science.gov (United States)

    Kirschbaum, Andreas; Kunz, Julia; Steinfeldt, Thorsten; Pehl, Anika; Meyer, Christian; Bartsch, Detlef K

    2014-12-01

    In every anatomic lung resection operation, the pulmonary artery itself or its branches must be sealed. This involves either stapling or ligating the vessels. Based on the positive results with the bipolar vessel sealing ≤7 mm in abdominal surgery the present study aimed to evaluate burst pressures of the pulmonary artery after sealing with the sealing instrument SealSafe G3 (Gebrüder Martin & CoKG, Tuttlingen, Germany). The whole pulmonary artery above the pulmonary valve was exposed up to the periphery of the left lung in freshly removed pig heart-lung blocks. A pressure-measuring cylinder was then implanted in the prepared vessel on the side at the main trunk of the pulmonary artery to determine the pressure in the vessel. After either ligation or bipolar sealing of the pulmonary artery, the pneumatic burst pressure (millimeters of mercury) was determined in a water bath. Three groups (n = 12 for each seal type) with different vessel diameters were examined: group 1: 0-6 mm, group 2: 7-12 mm, and group 3: >12 mm. In all cases, vessel sealing was performed with a MARSEAL 5 instrument (Gebrüder Martin & Co KG, Tuttlingen, Germany) and the SealSafe G3 current. The mean burst pressures of the individual groups (ligature and bipolar sealing) were compared using two-tailed, nonparametric Mann-Whitney U test. Significance was defined as P < 0.05. The mean burst pressures in group 1 were measured by 340 ± 13.4 mm Hg with ligature and 205 ± 44.4 mm Hg with bipolar sealing (P < 0.001). In group 2, the mean values obtained were 270 ± 28.2 mm Hg for ligature and 162 ± 36.0 mm Hg for bipolar sealing (P < 0.001). In group 3, the mean burst pressures for bipolar sealing were only 52.1 ± 15.1 mm Hg, whereas those for ligated vessels were 253 ± 46.9 mm Hg (P < 0.001). For this size of vessel the burst pressure was also determined after stapling. The mean value in this case was 230 ± 21.8 mm Hg. In all groups, the mean burst

  14. Recent investigations and tests with the BBR winding system for circumferential prestressing of concrete vessels and containments

    International Nuclear Information System (INIS)

    Schuett, K.; Speck, F.E.

    1993-01-01

    Prestressed concrete pressure vessels for nuclear power stations need post-tensioning systems of large capacity. For the circumferential prestressing, the continuous winding of prestressing steel has several advantages when compared to the use of large numbers of single tendons. About 15 years ago Bureau BBR Ltd (Zuerich) developed the winding system SW 8500. The further development work interrupted at that time for lack of immediate applications was resumed 4 years ago by Bureau BBR together with SUSPA on the ground of new projects being evaluated

  15. Computational hydrodynamic comparison of a mini vessel and a USP 2 dissolution testing system to predict the dynamic operating conditions for similarity of dissolution performance.

    Science.gov (United States)

    Wang, Bing; Bredael, Gerard; Armenante, Piero M

    2018-03-25

    The hydrodynamic characteristics of a mini vessel and a USP 2 dissolution testing system were obtained and compared to predict the tablet-liquid mass transfer coefficient from velocity distributions near the tablet and establish the dynamic operating conditions under which dissolution in mini vessels could be conducted to generate concentration profiles similar to those in the USP 2. Velocity profiles were obtained experimentally using Particle Image Velocimetry (PIV). Computational Fluid Dynamics (CFD) was used to predict the velocity distribution and strain rate around a model tablet. A CFD-based mass transfer model was also developed. When plotted against strain rate, the predicted tablet-liquid mass transfer coefficient was found to be independent of the system where it was obtained, implying that a tablet would dissolve at the same rate in both systems provided that the concentration gradient between the tablet surface and the bulk is the same, the tablet surface area per unit liquid volume is identical, and the two systems are operated at the appropriate agitation speeds specified in this work. The results of this work will help dissolution scientists operate mini vessels so as to predict the dissolution profiles in the USP 2, especially during the early stages of drug development. Copyright © 2018 Elsevier B.V. All rights reserved.

  16. Development of ultrasonic testing technique with the large transducer to inspect the containment vessel plates of nuclear power plant embedded in concrete

    International Nuclear Information System (INIS)

    Ishida, Hitoshi; Kurozumi, Yasuo; Kaneshima, Yoshiari

    2004-01-01

    The containment vessel plates embedded in concrete on Pressurized Water Reactors are inaccessible to inspect directly. Therefore, it is advisable to prepare inspection technology to detect existence and a location of corrosion on the embedded plates indirectly. In order to establish ultrasonic testing technique to be able to inspect the containment vessel plates embedded in concrete widely at the accessible point, experiments to detect artificial hollows simulating corrosion on a surface of a carbon steel plate mock-up covered with concrete simulating the embedded containment vessel plates were carried out with newly made ultrasonic transducers. We made newly low frequency (0.3 MHz and 0.5 MHz) surface shear horizontal (SH) wave transducers combined with three large active elements, which were equivalent to a 120mm width element. As a result of the experiments, the surface SH transducers could detect clearly the echo from the hollows with a depth of 9.5 mm and 19 mm at a distance of 1500mm from the transducers on the surface of the mock-up covered with concrete. Therefore, we evaluate that it is possible to detect the defects such as corrosion on the plates embedded in concrete with the newly made low frequency surface SH transducers with large elements. (author)

  17. Capabilities of the Power Burst Facility

    International Nuclear Information System (INIS)

    Spencer, W.A.; Jensen, A.M.; McCardell, R.K.

    1982-01-01

    The unique and diverse test capabilities of the Power Burst Facility (PBF) are described in this paper. The PBF test reactor, located at the Idaho National Engineering Laboratory, simulates normal, off-normal, and accident operating conditions of light water reactor fuel rods. An overview description is given, with specific detail on design and operating characteristics of the driver core, experiment test loop, fission product detection system, test train assembly facility, and support equipment which make the testing capability of the PBF so versatile

  18. Containment vessel

    International Nuclear Information System (INIS)

    Zbirohowski-Koscia, K.F.; Roberts, A.C.

    1980-01-01

    A concrete containment vessel for nuclear reactors is disclosed that is spherical and that has prestressing tendons disposed in first, second and third sets, the tendons of each set being all substantially concentric and centred around a respective one of the three orthogonal axes of the sphere; the tendons of the first set being anchored at each end at a first anchor rib running around a circumference of the vessel, the tendons of the second set being anchored at each end at a second anchor rib running around a circumference of the sphere and disposed at 90 0 to the first rib, and the tendons of the third set being anchored some to the first rib and the remainder to the second rib. (author)

  19. Quantum key based burst confidentiality in optical burst switched networks.

    Science.gov (United States)

    Balamurugan, A M; Sivasubramanian, A

    2014-01-01

    The optical burst switching (OBS) is an emergent result to the technology concern that could achieve a feasible network in future. They are endowed with the ability to meet the bandwidth requirement of those applications that require intensive bandwidth. There are more domains opening up in the OBS that evidently shows their advantages and their capability to face the future network traffic. However, the concept of OBS is still far from perfection facing issues in case of security threat. The transfer of optical switching paradigm to optical burst switching faces serious downfall in the fields of burst aggregation, routing, authentication, dispute resolution, and quality of service (QoS). This paper deals with employing RC4 (stream cipher) to encrypt and decrypt bursts thereby ensuring the confidentiality of the burst. Although the use of AES algorithm has already been proposed for the same issue, by contrasting the two algorithms under the parameters of burst encryption and decryption time, end-to-end delay, it was found that RC4 provided better results. This paper looks to provide a better solution for the confidentiality of the burst in OBS networks.

  20. Quantum Key Based Burst Confidentiality in Optical Burst Switched Networks

    Directory of Open Access Journals (Sweden)

    A. M. Balamurugan

    2014-01-01

    Full Text Available The optical burst switching (OBS is an emergent result to the technology concern that could achieve a feasible network in future. They are endowed with the ability to meet the bandwidth requirement of those applications that require intensive bandwidth. There are more domains opening up in the OBS that evidently shows their advantages and their capability to face the future network traffic. However, the concept of OBS is still far from perfection facing issues in case of security threat. The transfer of optical switching paradigm to optical burst switching faces serious downfall in the fields of burst aggregation, routing, authentication, dispute resolution, and quality of service (QoS. This paper deals with employing RC4 (stream cipher to encrypt and decrypt bursts thereby ensuring the confidentiality of the burst. Although the use of AES algorithm has already been proposed for the same issue, by contrasting the two algorithms under the parameters of burst encryption and decryption time, end-to-end delay, it was found that RC4 provided better results. This paper looks to provide a better solution for the confidentiality of the burst in OBS networks.

  1. Sensitivity tests on the rates of the excited states of positron decays during the rapid proton capture process of the one-zone X-ray burst model

    Science.gov (United States)

    Lau, Rita

    2018-02-01

    In this paper, we investigate the sensitivities of positron decays on a one-zone model of type-I X-ray bursts. Most existing studies have multiplied or divided entire beta decay rates (electron captures and beta decay rates) by 10. Instead of using the standard Fuller & Fowler (FFNU) rates, we used the most recently developed weak library rates [1], which include rates from Langanke et al.'s table (the LMP table) (2000) [2], Langanke et al.'s table (the LMSH table) (2003) [3], and Oda et al.'s table (1994) [4] (all shell model rates). We then compared these table rates with the old FFNU rates [5] to study differences within the final abundances. Both positron decays and electron capture rates were included in the tables. We also used pn-QRPA rates [6,7] to study the differences within the final abundances. Many of the positron rates from the nuclei's ground states and initial excited energy states along the rapid proton capture (rp) process have been measured in existing studies. However, because temperature affects the rates of excited states, these studies should have also acknowledged the half-lives of the nuclei's excited states. Thus, instead of multiplying or dividing entire rates by 10, we studied how the half-lives of sensitive nuclei in excited states affected the abundances by dividing the half-lives of the ground states by 10, which allowed us to set the half-lives of the excited states. Interestingly, we found that the peak of the final abundance shifted when we modified the rates from the excited states of the 105Sn positron decay rates. Furthermore, the abundance of 80Zr also changed due to usage of pn-QRPA rates instead of weak library rates (the shell model rates).

  2. Flaw preparations for HSST program vessel fracture mechanics testing: mechanical-cyclic pumping and electron-beam weld-hydrogen-charge cracking schemes

    International Nuclear Information System (INIS)

    Holz, P.P.

    1980-06-01

    The purpose of the document is to present schemes for flaw preparations in heavy section steel. The ability of investigators to grow representative sharp cracks of known size, location, and orientation is basic to representative field testing to determine data for potential flaw propagation, fracture behavior, and margin against fracture for high-pressure-, high-temperature-service steel vessels subjected to increasing pressurization and/or thermal shock. Gaging for analytical stress and strain procedures and ultrasonic and acoustic emission instrumentation can then be applied to monitor the vessel during testing and to study crack growth. This report presents flaw preparations for HSST fracture mechanics testing. Cracks were grown by two techniques: (1) a mechanical method wherein a premachined notch was sharpened by pressurization and (2) a method combining electron-beam welds and hydrogen charging to crack the chill zone of a rapidly placed autogenous weld. The mechanical method produces a naturally occurring growth shape controlled primarily by the shape of the machined notch; the welding-electrochemical method produces flaws of uniform depth from the surface of a wall or machined notch. Theories, details, discussions, and procedures are covered for both of the flaw-growing schemes

  3. In-vessel natural circulation during a hypothetical loss-of-heat-sink accident in the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Perkins, K.R.; Bari, R.A.; Pratt, W.T.

    1979-05-01

    The capability to remove decay heat from the FFTF core via in-vessel natural circulation has been analyzed for the preboiling phase using a lumped parameter model. The results indicate that boiling will occur in the average fuel assembly for a wide spectrum of initial conditions which appear to be representative of the hypothetical loss-of-heat-sink accident. Two-phase pressure drop calculations indicate that, once the saturation temperature is reached, coolability can only be assured for decay heat levels which are less than 0.5% of the operating power. A review of the limited sodium boiling data indicates that boiling-induced natural circulation may support up to 4% of the operating power, but geometric atypicalities and a large degree of inlet subcooling for the existing data limit the applicability to the loss-of-heat-sink accident in FFTF

  4. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    International Nuclear Information System (INIS)

    Nakahira, M.; Takeda, N.; Kakudate, S.; Onozuka, M.

    2008-01-01

    The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV

  5. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, M. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan)], E-mail: nakahira.masataka@jaea.go.jp; Takeda, N.; Kakudate, S. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Onozuka, M. [Mitsubishi Nuclear Energy Systems, Inc., 1700K Street NW, Suite 440, Washington, DC 20006 (United States)

    2008-12-15

    The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV.

  6. Structural model testing for prestressed concrete pressure vessels: a study of grouted vs nongrouted posttensioned prestressing tendon systems

    International Nuclear Information System (INIS)

    Naus, D.J.

    1979-04-01

    Nongrouted tendons are predominantly used in this country as the prestressing system for prestressed concrete pressure vessels (PCPVs) because they are more easily surveyed to detect reductions in prestressing level and distress such as results from corrosion. Grouted tendon systems, however, offer advantages which may make them cost-effective for PCPV applications. Literature was reviewed to (1) provide insight on the behavior of grouted tendon system, (2) establish performance histories for structures utilizing grouted tendons, (3) examine corrosion protection procedures for prestressing tendons, (4) identify arguments for and against using grouted tendons, and (5) aid in the development of the experimental investigation. The experimental investigation was divided into four phases: (1) grouted-nongrouted tendon behavior, (2) evaluation of selected new material systems, (3) bench-scale corrosion studies, and (4) preliminary evaluation of acoustic emission techniques for monitoring grouted tendons in PCPVs. The groutability of large tendon systems was also investigated

  7. Definition of the minimum longitude of insert in the rebuilding of Charpy test tubes for surveillance and life extension of vessels in Mexico

    International Nuclear Information System (INIS)

    Romero C, J.; Hernandez C, R.; Rocamontes A, M.

    2011-11-01

    In the National Institute of Nuclear Research (Mexico) a welding system for the rebuilding of Charpy test tubes has been developed, automated, qualified and used for the surveillance of the mechanical properties (mainly embrittlement) of the vessel. This system uses the halves of the rehearsed Charpy test tubes of the surveillance capsules extracted of the reactors, to obtain, of a rehearsed test tube, two reconstituted test tubes. This rebuilding process is used so much in the surveillance program like in the potential extension of the operation license of the vessel. To the halves of Charpy test tubes that have been removed the deformed part by machine are called -insert- and in a very general way the rebuilding consists in weld with the welding process -Stud Welding- two metallic implants in the ends of the insert, to obtain a reconstituted test tube. The main characteristic of this welding are the achieved small dimensions, so much of the areas welded as of the areas affected by the heat. The applicable normative settles down that the minim longitude of the insert for the welding process by Stud Welding it should be of 18 mm, however according to the same normative this longitude can diminish if is demonstrated analytic or experimentally that the central volume of 1 cm 3 in the insert is not affected. In this work the measurement of the temperature profiles to different distances of the welding interface is presented, defining an equation for the maximum temperatures reached in function of the distance, on the other hand the real longitude affected in the test tube by means of metallography is determined and this way the minimum longitude of the insert for this developed rebuilding system was determined. (Author)

  8. Rapid sealing of porcine renal blood vessels, ex vivo, using a high power, 1470-nm laser, and laparoscopic prototype

    Science.gov (United States)

    Hardy, Luke A.; Hutchens, Thomas C.; Larson, Eric R.; Gonzalez, David A.; Chang, Chun-Hung; Nau, William H.; Fried, Nathaniel M.

    2017-05-01

    Energy-based, radiofrequency (RF) and ultrasonic (US) devices currently provide rapid sealing of blood vessels during laparoscopic procedures. We are exploring infrared lasers as an alternate energy modality for vessel sealing, capable of generating less collateral thermal damage. Previous studies demonstrated feasibility of sealing vessels in an in vivo porcine model using a 1470-nm laser. However, the initial prototype was designed for testing in open surgery and featured tissue clasping and light delivery mechanisms incompatible with laparoscopic surgery. In this study, a laparoscopic prototype similar to devices currently in surgical use was developed, and performance tests were conducted on porcine renal blood vessels, ex vivo. The 5-mm outer-diameter laparoscopic prototype featured a traditional Maryland jaw configuration that enables tissue manipulation and blunt dissection. Laser energy was delivered through a 550-μm-core-diameter optical fiber with side-delivery from the lower jaw and beam dimensions of 18-mm length×1.2-mm width. The 1470-nm diode laser delivered 68 W with 3-s activation time, consistent with vessel seal times associated with RF and US-based devices. A total of 69 fresh porcine renal vessels with mean diameter of 3.3±1.7 mm were tested, ex vivo. Vessels smaller than 5-mm diameter were consistently sealed (48/51) with burst pressures greater than malignant hypertension blood pressure (180 mmHg), averaging 1038±474 mmHg. Vessels larger than 5 mm were not consistently sealed (6/18), yielding burst pressures of only 174±221 mmHg. Seal width, thermal damage zone, and thermal spread averaged 1.7±0.8, 3.4±0.7, and 1.0±0.4 mm, respectively. Results demonstrated that the 5-mm optical laparoscopic prototype consistently sealed vessels less than 5-mm diameter with low thermal spread. Further in vivo studies are planned to test the performance across a variety of vessels and tissues.

  9. Development of a sealing process of capsules for surveillance test tubes of the vessel in nuclear power plants

    International Nuclear Information System (INIS)

    Romero C, J.; Fernandez T, F.; Perez R, N.; Rocamontes A, M.; Garcia R, R.

    2007-01-01

    The surveillance capsule is composed by the support, three capsules for impact test tubes, five capsules for tension test tubes and one porta dosemeters. The capsules for test tubes are of two types: rectangular capsule for Charpy test tubes and cylindrical capsule for tension test tubes. This work describes the development of the welding system to seal the capsules for test tubes that should contain helium of ultra high purity to a pressure of 1 atmosphere. (Author)

  10. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    Energy Technology Data Exchange (ETDEWEB)

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  11. Solar microwave bursts - A review

    Science.gov (United States)

    Kundu, M. R.; Vlahos, L.

    1982-01-01

    Observational and theoretical results on the physics of microwave bursts that occur in the solar atmosphere are reviewed. Special attention is given to the advances made in burst physics over the last few years with the great improvement in spatial and time resolution, especially with instruments like the NRAO three-element interferometer, the Westerbork Synthesis Radio Telescope, and more recently the Very Large Array. Observations made on the preflare build-up of an active region at centimeter wavelengths are reviewed. Three distinct phases in the evolution of cm bursts, namely the impulsive phase, the post-burst phase, and the gradual rise and fall, are discussed. Attention is also given to the flux density spectra of centimeter bursts. Descriptions are given of observations of fine structures with temporal resolution of 10-100 ms in the intensity profiles of cm-wavelength bursts. High spatial resolution observations are analyzed, with special reference to the one- and two-dimensional maps of cm burst sources.

  12. Apparent spatial uniformity of the gamma-ray bursts detected by the Konus experiment on Venera 11 and Venera 12

    International Nuclear Information System (INIS)

    Higdon, J.C.; Schmidt, M.

    1990-01-01

    The V/Vmax test is applied to gamma-ray bursts of duration longer than 1 sec recorded by the Konus experiment, to examine quantitatively the uniformity of the burst source population. A sample of 123 bursts detected on Venera 11 and Venera 12, gives mean V/Vmax = 0.45 + or - 0.03, consistent with 0.5, the value expected for a uniform distribution in space of the parent population of burst sources. It is argued that experimenters give careful attention to the detection limit for each recorded gamma-ray burst, and that quantitative data for burst properties and detection limits should be published. 28 refs

  13. High repetition rate burst-mode spark gap

    International Nuclear Information System (INIS)

    Faltens, A.; Reginato, L.; Hester, R.; Chesterman, A.; Cook, E.; Yokota, T.; Dexter, W.

    1978-01-01

    Results are presented on the design and testing of a pressurized gas blown spark gap switch capable of high repetition rates in a burst mode of operation. The switch parameters which have been achieved are as follows: 220-kV, 42-kA, a five pulse burst at 1-kHz, 12-ns risetime, 2-ns jitter at a pulse width of 50-ns

  14. Return to normal of sup(99m)Tc-plasmin test after deep venous thrombosis and its relationship to vessel wall fibrinolysis

    Energy Technology Data Exchange (ETDEWEB)

    Edenbrandt, C.M.; Hedner, U.; Tengborn, L.; Nilsson, J.; Ohlin, P.

    1986-08-01

    Fourteen patients with deep venous thrombosis (DVT) and a positive sup(99m)Tc-plasmin test were followed up to determine how soon a negative test was obtained. Localization and extension of the thrombi were determined by phlebography. Plasminogen activator activity in vein walls and local fibrinolytic activity after venous occlusion were measured in order to find out what the prerequisites for impaired thrombolysis are. The time required to obtain a negative sup(99m)Tc-plasmin test showed considerable variation, ranging from less than 1 week to more than 6 months. The sup(99m)Tc-plasmin test had returned to normal in 64% of the patients after 6 months. No relationship was found between vessel wall fibrinolysis and time to normalization. Instead, we found an association between the time to normalization of the sup(99m)Tc-plasmin test and the size of the thrombus, according to phlebography, as well as between the time to normalization of the sup(99m)Tc-plasmin test and the extension of leg points with a positive sup(99m)Tc-plasmin test at admission. The finding of abnormal sup(99m)Tc-plasmin test results more than 6 months after acute DVT is of practical importance and warrants caution when evaluating patients with symptoms and signs suggestive of acute recurrent DVT.

  15. Solar X-ray bursts

    International Nuclear Information System (INIS)

    Urnov, A.M.

    1980-01-01

    In the popular form the consideration is given to the modern state tasks and results of X-ray spectrometry of solar bursts. The operation of X-ray spectroheliograph is described. Results of spectral and polarization measurings of X-ray radiation of one powerful solar burst are presented. The conclusion has been drawn that in the process of burst development three characteristic stages may be distingwished: 1) the initial phase; just in this period processes which lead to observed consequences-electromagnetic and corpuscular radiation are born; 2) the impulse phase, or the phase of maximum, is characterised by sharp increase of radiation flux. During this phase the main energy content emanates and some volumes of plasma warm up to high temperatures; 3) the phase of burst damping, during which plasma cools and reverts to the initial condition

  16. Numerical simulation of a Charpy test and correlation of fracture toughness with fracture energy. Vessel steel and duplex stainless steel of the primary loop

    International Nuclear Information System (INIS)

    Breban, P; Eripret, C.

    1995-01-01

    The analysis methods used to evaluate the harmlessness of defects in the components of the primary coolant circuit of pressurized water reactor are based on the knowledge of the failure properties of concerned materials. The toughness is used to be measured through tests performed on normalized samples. But in some cases, especially for the vessel steel submitted to irradiation effects or for cast components in duplex stainless steel sensitive to thermal ageing, these measurements are not available on the material aged in operation. Therefore, fracture resistance has been evaluated through Charpy tests. Toughness is thus obtained on the basis of an empirical correlation. To improve these predictions, a modeling of the Charpy test in the framework of the local approach to fracture has been performed, for both materials. For the vessel steel, a complete evaluation of toughness has been achieved on the basis of a bidimensional viscoplastic modeling under large strain assumptions and a post-treatment with a Weibull model (cleavage fracture). The main hypothesis (partition between plain stress and plain strain areas in the bidimensional modeling) was corrected after a three dimensional calculations with the finite element program Code-Aster. The fracture analysis put into evidence that damage considerations like cavity nucleation and growth have to be introduced in the model in order to improve the description of physical phenomena. Two ways of progress have been suggested and are in course of being investigated, one in the framework of local approach to failure, the other with the help of micro-macro relationship. With regard to the duplex steel, the description of a Charpy (U) test allowed to clearly discriminate between crack initiation and propagation phases. A modeling through an equivalent homogenous material with a damage law based on a modified Gurson potential enables to describe quantitatively both phases of fracture. It clearly appears that a reliable

  17. Prediction of failure strain and burst pressure in high yield-to-tensile strength ratio linepipe

    International Nuclear Information System (INIS)

    Law, M.; Bowie, G.

    2007-01-01

    Failure pressures and strains were predicted for a number of burst tests as part of a project to explore failure strain in high yield-to-tensile strength ratio linepipe. Twenty-three methods for predicting the burst pressure and six methods of predicting the failure strain are compared with test results. Several methods were identified which gave accurate and reliable estimates of burst pressure. No method of accurately predicting the failure strain was found, though the best was noted

  18. Prediction of failure strain and burst pressure in high yield-to-tensile strength ratio linepipe

    Energy Technology Data Exchange (ETDEWEB)

    Law, M. [Institute of Materials and Engineering Science, Australian Nuclear Science and Technology Organisation (ANSTO), Lucas Heights, NSW (Australia)]. E-mail: mlx@ansto.gov.au; Bowie, G. [BlueScope Steel Ltd., Level 11, 120 Collins St, Melbourne, Victoria 3000 (Australia)

    2007-08-15

    Failure pressures and strains were predicted for a number of burst tests as part of a project to explore failure strain in high yield-to-tensile strength ratio linepipe. Twenty-three methods for predicting the burst pressure and six methods of predicting the failure strain are compared with test results. Several methods were identified which gave accurate and reliable estimates of burst pressure. No method of accurately predicting the failure strain was found, though the best was noted.

  19. 30 CFR 57.3461 - Rock bursts.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Rock bursts. 57.3461 Section 57.3461 Mineral...-Underground Only § 57.3461 Rock bursts. (a) Operators of mines which have experienced a rock burst shall— (1) Within twenty four hours report to the nearest MSHA office each rock burst which: (i) Causes persons to...

  20. Chimera states in bursting neurons

    OpenAIRE

    Bera, Bidesh K.; Ghosh, Dibakar; Lakshmanan, M.

    2015-01-01

    We study the existence of chimera states in pulse-coupled networks of bursting Hindmarsh-Rose neurons with nonlocal, global and local (nearest neighbor) couplings. Through a linear stability analysis, we discuss the behavior of stability function in the incoherent (i.e. disorder), coherent, chimera and multi-chimera states. Surprisingly, we find that chimera and multi-chimera states occur even using local nearest neighbor interaction in a network of identical bursting neurons alone. This is i...

  1. Detection circuit for gamma-ray burst

    International Nuclear Information System (INIS)

    Murakami, Hiroyuki; Yamagami, Takamasa; Mori, Kunishiro; Uchiyama, Sadayuki.

    1982-01-01

    A new gamma-ray burst detection system is described. The system was developed as an environmental monitor of an accelerator, and can be used as the burst detection system. The system detects the arrival time of burst. The difference between the arrival times detected at different places will give information on the burst source. The frequency of detecting false burst was estimated, and the detection limit under the estimated frequency of false burst was also calculated. Decision whether the signal is false or true burst was made by the statistical treatment. (Kato, T.)

  2. Gamma-ray burst models.

    Science.gov (United States)

    King, Andrew

    2007-05-15

    I consider various possibilities for making gamma-ray bursts, particularly from close binaries. In addition to the much-studied neutron star+neutron star and black hole+neutron star cases usually considered good candidates for short-duration bursts, there are also other possibilities. In particular, neutron star+massive white dwarf has several desirable features. These systems are likely to produce long-duration gamma-ray bursts (GRBs), in some cases definitely without an accompanying supernova, as observed recently. This class of burst would have a strong correlation with star formation and occur close to the host galaxy. However, rare members of the class need not be near star-forming regions and could have any type of host galaxy. Thus, a long-duration burst far from any star-forming region would also be a signature of this class. Estimates based on the existence of a known progenitor suggest that this type of GRB may be quite common, in agreement with the fact that the absence of a supernova can only be established in nearby bursts.

  3. Solar Drift-Pair Bursts

    Science.gov (United States)

    Stanislavsky, A.; Volvach, Ya.; Konovalenko, A.; Koval, A.

    2017-08-01

    In this paper a new sight on the study of solar bursts historically called drift pairs (DPs) is presented. Having a simple morphology on dynamic spectra of radio records (two short components separated in time, and often they are very similar) and discovered at the dawn of radio astronomy, their features remain unexplained totally up to now. Generally, the DPs are observed during the solar storms of type III bursts, but not every storm of type III bursts is linked with DPs. Detected by ground-based instruments at decameter and meter wavelengths, the DP bursts are limited in frequency bandwidth. They can drift from high frequencies to low ones and vice versa. Their frequency drift rate may be both lower and higher than typical rates of type III bursts at the same frequency range. The development of low-frequency radio telescopes and data processing provide additional possibilities in the research. In this context the fresh analysis of DPs, made from recent observations in the summer campaign of 2015, are just considered. Their study was implemented by updated tools of the UTR-2 radio telescope at 9-33 MHz. During 10-12 July of 2015, DPs forming the longest patterns on dynamic spectra are about 7% of the total number of recorded DPs. Their marvelous resemblance in frequency drift rates with the solar S-bursts is discussed.

  4. Double-Ended Break Test of an 8.5 inch Direct Vessel Injection Line using the ATLAS

    International Nuclear Information System (INIS)

    Choi, Ki Yong; Kang, Kyoung Ho; Kim, Bok Deuk; Kim, Yeon Sik; Min, Kyoung Ho; Park, Choon Kyoung; Park, Hyun Sik; Baek, Won Pil; Cho, Seok; Choi, Nam Hyun

    2010-01-01

    A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at the Korea Atomic Energy Research Institute (KAERI). It is a 1/2 reduced height and 1/288-volume scaled test facility with respect to the APR1400, an evolutionary pressurized water reactor developed by the Korean industry. In 2008, an integral effect test for simulating a guillotine break of a DVI line of the APR1400 was carried out as the first DVI test item, named as SB-DVI-03, on May, 2008. With an improvement on the break flow measuring system, the second DVI test for a guillotine break of a DVI line of the APR1400, named as SB-DVI-08, was conducted for repeatability. The present data is the first integral effect test data of its kind for simulating a DVI line break accident. It will help in understanding the thermal hydraulic phenomena occurring during the DVI line break accident. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. The present integral effect test data will be used to validate the current safety analysis methodology for the DVI line break accident

  5. X-ray bursts: Observation versus theory

    Science.gov (United States)

    Lewin, W. H. G.

    1981-01-01

    Results of various observations of common type I X-ray bursts are discussed with respect to the theory of thermonuclear flashes in the surface layers of accreting neutron stars. Topics covered include burst profiles; irregular burst intervals; rise and decay times and the role of hydrogen; the accuracy of source distances; accuracy in radii determination; radius increase early in the burst; the super Eddington limit; temperatures at burst maximum; and the role of the magnetic field.

  6. Bursting pressure of autofrettaged cylinders with inclined external cracks

    International Nuclear Information System (INIS)

    Seifi, Rahman; Babalhavaeji, Majid

    2012-01-01

    Autofrettaging a pressure vessel improves its pressure capacity. This is reliable if there isn’t any crack or other type of flaws. In this paper, the effects of external surface cracks on bursting pressure of autofrettaged cylinders are studied. It is observed that bursting pressure decreases considerably (up to 30%) due to external cracks in the cylinders without autofrettage. This reduction increases for high levels of the applied autofrettage. External axial cracks have more effects than inclined cracks. Comparing experimental and numerical results show that the numerical methods can acceptably predict the bursting pressure of the autofrettaged cracked cylinders. These predictions are valid when the fracture parameter (J-Integral) is calculated from the modified equation that takes into account the effects of residual stresses. - Highlights: ► Modified J-Integral can be used for study of autofrettaged cracked cylinders. ► External axial cracks reduce considerably the pressure capacity of cylinders. ► External circumferential cracks have not considerable effects on bursting pressure. ► Autofrettage has contrary effects on external crack in compared with internal crack.

  7. Bursting pressure of autofrettaged cylinders with inclined external cracks

    Energy Technology Data Exchange (ETDEWEB)

    Seifi, Rahman, E-mail: rseifi@basu.ac.ir [Mechanical Engineering Department, Faculty of Engineering, Bu-Ali Sina University, Hamedan (Iran, Islamic Republic of); Babalhavaeji, Majid [Mechanical Engineering Department, Faculty of Engineering, Bu-Ali Sina University, Hamedan (Iran, Islamic Republic of)

    2012-01-15

    Autofrettaging a pressure vessel improves its pressure capacity. This is reliable if there isn't any crack or other type of flaws. In this paper, the effects of external surface cracks on bursting pressure of autofrettaged cylinders are studied. It is observed that bursting pressure decreases considerably (up to 30%) due to external cracks in the cylinders without autofrettage. This reduction increases for high levels of the applied autofrettage. External axial cracks have more effects than inclined cracks. Comparing experimental and numerical results show that the numerical methods can acceptably predict the bursting pressure of the autofrettaged cracked cylinders. These predictions are valid when the fracture parameter (J-Integral) is calculated from the modified equation that takes into account the effects of residual stresses. - Highlights: Black-Right-Pointing-Pointer Modified J-Integral can be used for study of autofrettaged cracked cylinders. Black-Right-Pointing-Pointer External axial cracks reduce considerably the pressure capacity of cylinders. Black-Right-Pointing-Pointer External circumferential cracks have not considerable effects on bursting pressure. Black-Right-Pointing-Pointer Autofrettage has contrary effects on external crack in compared with internal crack.

  8. An experimental study on feasibility of ex-vessel cooling through the external guide vessel

    International Nuclear Information System (INIS)

    Kang, Kyoung-Ho; Kim, Jong-Hwan; Park, Rae-Jun; Kim, Sang-Baik

    2000-01-01

    This paper presents the results of a series of experiments for assessing the efficacy of ex-vessel cooling through the external guide vessel during a severe accident. Four tests were performed in the LAVA test facility at KAERI, varying the boundary conditions at the outer surface of the vessel. The first test was a dry condition test conducted without cooling the outside of the vessel. On the other hand, in the second test, the cooling of the vessel surface was produced by gravity-driven forced injection of water along the annular gap of 25 mm between the vessel and the external guide vessel. Water flow rate was about 0.85 kg/s and total mass of available water was 300 kg. For the evaluation of the water flow rate effect, the third test was performed with a pool type cooling in the annulus without any circulation of water. These two external cooling tests were performed under elevated pressure of about 1.6 MPa. Finally, the fourth test was conducted under atmospheric pressure to evaluate the effect of system pressure on boiling heat transfer characteristics. In the dry test and the pool type ex-vessel cooling test performed under atmospheric pressure, the vessel was failed by a melt penetration at about 40 degree upper position from the vessel bottom, which is coincident with the boundary of the Al 2 O 3 /Fe melt separated layers. On the other hand, in both of the ex-vessel cooling tests conducted under elevated pressure of about 1.6 MPa, the vessel didn't fail. Compared with the pool boiling test, the vessel experienced effective cooling due to the inlet flow in the forced flow test. Synthesized the results of the tests, it was shown that the heat removal with ex-vessel cooling through the guide vessel is feasible, but the additional evaluations should be performed to guarantee enough thermal margin. (author)

  9. Nanostructural evolution in surveillance test specimens of a commercial nuclear reactor pressure vessel studied by three-dimensional atom probe and positron annihilation

    International Nuclear Information System (INIS)

    Toyama, T.; Nagai, Y.; Tang, Z.; Hasegawa, M.; Almazouzi, A.; Walle, E. van; Gerard, R.

    2007-01-01

    The nanostructural evolution of irradiation-induced Cu-rich nanoprecipitates (CRNPs) and vacancy clusters in surveillance test specimens of in-service commercial nuclear reactor pressure vessel steel welds of Doel-1 and Doel-2 are revealed by combining the three-dimensional local electrode atom probe and positron annihilation techniques. In both medium (0.13 wt.%) and high (0.30 wt.%) Cu welds, the CRNPs are found to form readily at the very beginning of the reactor lifetime. Thereafter, during the subsequent 30 years of operation, the residual Cu concentration in the matrix shows a slight decrease while the CRNPs coarsen. On the other hand, small vacancy clusters of V 3 -V 4 start appearing after the initial Cu precipitation and accumulate steadily with increasing neutron dose. The observed nanostructural evolution is shown to provide unique and fundamental information about the mechanisms of the irradiation-induced embrittlement of these specific materials

  10. Development of the front end test stand and vessel for extraction and source plasma analyses negative hydrogen ion sources at the Rutherford Appleton Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Lawrie, S. R., E-mail: scott.lawrie@stfc.ac.uk [STFC ISIS Pulsed Spallation Neutron and Muon Facility, Rutherford Appleton Laboratory, Harwell Oxford, Harwell (United Kingdom); John Adams Institute of Accelerator Science, University of Oxford, Oxford (United Kingdom); Faircloth, D. C.; Letchford, A. P.; Perkins, M.; Whitehead, M. O.; Wood, T. [STFC ISIS Pulsed Spallation Neutron and Muon Facility, Rutherford Appleton Laboratory, Harwell Oxford, Harwell (United Kingdom); Gabor, C. [ASTeC Intense Beams Group, Rutherford Appleton Laboratory, Harwell Oxford, Harwell (United Kingdom); Back, J. [High Energy Physics Department, University of Warwick, Coventry (United Kingdom)

    2014-02-15

    The ISIS pulsed spallation neutron and muon facility at the Rutherford Appleton Laboratory (RAL) in the UK uses a Penning surface plasma negative hydrogen ion source. Upgrade options for the ISIS accelerator system demand a higher current, lower emittance beam with longer pulse lengths from the injector. The Front End Test Stand is being constructed at RAL to meet the upgrade requirements using a modified ISIS ion source. A new 10% duty cycle 25 kV pulsed extraction power supply has been commissioned and the first meter of 3 MeV radio frequency quadrupole has been delivered. Simultaneously, a Vessel for Extraction and Source Plasma Analyses is under construction in a new laboratory at RAL. The detailed measurements of the plasma and extracted beam characteristics will allow a radical overhaul of the transport optics, potentially yielding a simpler source configuration with greater output and lifetime.

  11. Heterogeneity in Short Gamma-Ray Bursts

    Science.gov (United States)

    Norris, Jay P.; Gehrels Neil; Scargle, Jeffrey D.

    2011-01-01

    We analyze the Swift/BAT sample of short gamma-ray bursts, using an objective Bayesian Block procedure to extract temporal descriptors of the bursts' initial pulse complexes (IPCs). The sample comprises 12 and 41 bursts with and without extended emission (EE) components, respectively. IPCs of non-EE bursts are dominated by single pulse structures, while EE bursts tend to have two or more pulse structures. The medians of characteristic timescales - durations, pulse structure widths, and peak intervals - for EE bursts are factors of approx 2-3 longer than for non-EE bursts. A trend previously reported by Hakkila and colleagues unifying long and short bursts - the anti-correlation of pulse intensity and width - continues in the two short burst groups, with non-EE bursts extending to more intense, narrower pulses. In addition we find that preceding and succeeding pulse intensities are anti-correlated with pulse interval. We also examine the short burst X-ray afterglows as observed by the Swift/XRT. The median flux of the initial XRT detections for EE bursts (approx 6 X 10(exp -10) erg / sq cm/ s) is approx > 20 x brighter than for non-EE bursts, and the median X-ray afterglow duration for EE bursts (approx 60,000 s) is approx 30 x longer than for non-EE bursts. The tendency for EE bursts toward longer prompt-emission timescales and higher initial X-ray afterglow fluxes implies larger energy injections powering the afterglows. The longer-lasting X-ray afterglows of EE bursts may suggest that a significant fraction explode into more dense environments than non-EE bursts, or that the sometimes-dominant EE component efficiently p()wers the afterglow. Combined, these results favor different progenitors for EE and non-EE short bursts.

  12. HETEROGENEITY IN SHORT GAMMA-RAY BURSTS

    International Nuclear Information System (INIS)

    Norris, Jay P.; Gehrels, Neil; Scargle, Jeffrey D.

    2011-01-01

    We analyze the Swift/BAT sample of short gamma-ray bursts, using an objective Bayesian Block procedure to extract temporal descriptors of the bursts' initial pulse complexes (IPCs). The sample is comprised of 12 and 41 bursts with and without extended emission (EE) components, respectively. IPCs of non-EE bursts are dominated by single pulse structures, while EE bursts tend to have two or more pulse structures. The medians of characteristic timescales-durations, pulse structure widths, and peak intervals-for EE bursts are factors of ∼2-3 longer than for non-EE bursts. A trend previously reported by Hakkila and colleagues unifying long and short bursts-the anti-correlation of pulse intensity and width-continues in the two short burst groups, with non-EE bursts extending to more intense, narrower pulses. In addition, we find that preceding and succeeding pulse intensities are anti-correlated with pulse interval. We also examine the short burst X-ray afterglows as observed by the Swift/X-Ray Telescope (XRT). The median flux of the initial XRT detections for EE bursts (∼6x10 -10 erg cm -2 s -1 ) is ∼>20x brighter than for non-EE bursts, and the median X-ray afterglow duration for EE bursts (∼60,000 s) is ∼30x longer than for non-EE bursts. The tendency for EE bursts toward longer prompt-emission timescales and higher initial X-ray afterglow fluxes implies larger energy injections powering the afterglows. The longer-lasting X-ray afterglows of EE bursts may suggest that a significant fraction explode into denser environments than non-EE bursts, or that the sometimes-dominant EE component efficiently powers the afterglow. Combined, these results favor different progenitors for EE and non-EE short bursts.

  13. Analysis of the surveillance test data on irradiation embrittlement of the reactor pressure vessel steels in LWRs

    International Nuclear Information System (INIS)

    Lee, Gyoeng Geun; Kwon, Jun Hyun

    2010-11-01

    The surveillance test data in Korean LWRs were analyzed from a viewpoint of materials science. TTS change with the neutron irradiation were compared to the model values of the RG1.99/2 and NUREG/CR-6551. The model values of TTS were higher than the actual values of TTS. It was impossible to find a relationship between TTS and neutron fluence in weld data. The correlation of the increase in YS (yield strength) and TTS with neutron irradiation was also investigated. Like the result of TTS change, the YS/TTS showed the correlations in plate/forgings metals, however no correlation in weld metals. The data were similar to Odette's result about US surveillance tests. From the empirical relationships, the TTS curve change could be predicted using the CVN test result of the unirradiated specimen and the change in YS with neutron irradiation of the specimen

  14. Computer-assisted acoustic emission analysis in alternating current magnetization and hardness testing of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Blochwitz, M.; Kretzschmar, F.; Rattke, R.

    1985-01-01

    Non-destructive determination of material characteristics such as nilductility transition temperature is of high importance in component monitoring during long-term operation. An attempt has been made to obtain characteristics correlating with mechanico-technological material characteristics by both acoustic resonance through magnetization (ARDM) and acoustic emission analysis in Vickers hardness tests. Taking into account the excitation mechanism of acoustic emission generation, which has a quasistationary stochastic character in a.c. magnetization and a transient nature in hardness testing, a microcomputerized device has been constructed for frequency analysis of the body sound level in ARDM evaluation and for measuring the pulse sum and/or pulse rate during indentation of the test specimen in hardness evaluation. Prerequisite for evaluating the measured values is the knowledge of the frequency dependence of the sensors and the instrument system. The results obtained are presented. (author)

  15. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

    1998-01-01

    Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use

  16. Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Van de Velde, J.; Fabry, A.; Van Walle, E.; Chaoudi, R

    1998-07-01

    SCK-CEN's R and D programme on Reactor Pressure Vessel (RPV) Steels in performed in support of the RVP integrity assessment. Its main objectives are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate the applied methodology on a broad database; (3) to achieve regulatory acceptance and industrial use. Progress and achievements in 1999 are reported.

  17. Assessment of MARS for downcomer multi-dimensional thermal hydraulics during LBLOCA reflood using KAERI air-water direct vessel injection tests

    Energy Technology Data Exchange (ETDEWEB)

    Won-Jae, Lee; Kwi-Seok, Ha; Chul-Hwa, Song [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)

    2001-07-01

    The MARS code has been assessed for the downcomer multi-dimensional thermal hydraulics during a large break loss-of-coolant accident (LBLOCA) reflood of Korean Next Generation Reactor (KNGR) that adopted an upper direct vessel injection (DVI) design. Direct DVI bypass and downcomer level sweep-out tests carried out at 1/50-scale air-water DVI test facility are simulated to examine the capability of MARS. Test conditions are selected such that they represent typical reflood conditions of KNGR, that is, DVI injection velocities of 1.0 {approx} 1.6 m/sec and air injection velocities of 18.0 {approx} 35.0 m/sec, for single and double DVI configurations. MARS calculation is first adjusted to the experimental DVI film distribution that largely affects air-water interaction in a scaled-down downcomer, then, the code is assessed for the selected test matrix. With some improvements of MARS thermal-hydraulic (T/H) models, it has been demonstrated that the MARS code is capable of simulating the direct DVI bypass and downcomer level sweep-out as well as the multi-dimensional thermal hydraulics in downcomer, where condensation effect is excluded. (authors)

  18. Temperature-dependent attenuation of ex-vessel flux measurements at the Hanford Fast Flux Test Facility

    International Nuclear Information System (INIS)

    McLane, F.E.; Wood, M.R.; Rathbun, J.L.

    1982-01-01

    Indicated nuclear power, developed by measuring leakage neutrons, has been found to be temperature dependent at the Hanford Fast Flux Test Facility (FFTF). The magnitude, sense and speed of response of the effect suggest that hot sodium above th core and shield is a significant cause. Future designs which may minimize this effect are discussed

  19. Research on friction coefficient of nuclear Reactor Vessel Internals Hold Down Spring: Stress coefficient test analysis method

    International Nuclear Information System (INIS)

    Linjun, Xie; Guohong, Xue; Ming, Zhang

    2016-01-01

    Graphical abstract: HDS stress coefficient test apparatus. - Highlights: • This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. • The mathematical relation between the load and the strain is obtained about the HDS, and the mathematical model of the stress coefficient and the friction coefficient is established. So, a set of test apparatuses for obtaining the stress coefficient is designed according to the model scaling criterion and the friction coefficient of the K1000 HDS is calculated to be 0.336 through the obtained stress coefficient. • The relation curve between the theoretical load and the friction coefficient is obtained through analysis and indicates that the change of the friction coefficient f would influence the pretightening load under the condition of designed stress. The necessary pretightening load in the design process is calculated to be 5469 kN according to the obtained friction coefficient. Therefore, the friction coefficient and the pretightening load under the design conditions can provide accurate pretightening data for the analysis and design of the reactor HDS according to the operations. - Abstract: This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. By carrying out tests and researches through a stress testing technique, P–σ curves in loading and unloading processes of the HDS are obtained and the stress coefficient k f of the HDS is obtained. So, the

  20. Research on friction coefficient of nuclear Reactor Vessel Internals Hold Down Spring: Stress coefficient test analysis method

    Energy Technology Data Exchange (ETDEWEB)

    Linjun, Xie, E-mail: linjunx@zjut.edu.cn [College of Mechanical Engineering, Zhejiang University of Technology, Hangzhou 310014 (China); Guohong, Xue; Ming, Zhang [Shanghai Nuclear Engineering Research & Design Institute, Shanghai 200233 (China)

    2016-08-01

    Graphical abstract: HDS stress coefficient test apparatus. - Highlights: • This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. • The mathematical relation between the load and the strain is obtained about the HDS, and the mathematical model of the stress coefficient and the friction coefficient is established. So, a set of test apparatuses for obtaining the stress coefficient is designed according to the model scaling criterion and the friction coefficient of the K1000 HDS is calculated to be 0.336 through the obtained stress coefficient. • The relation curve between the theoretical load and the friction coefficient is obtained through analysis and indicates that the change of the friction coefficient f would influence the pretightening load under the condition of designed stress. The necessary pretightening load in the design process is calculated to be 5469 kN according to the obtained friction coefficient. Therefore, the friction coefficient and the pretightening load under the design conditions can provide accurate pretightening data for the analysis and design of the reactor HDS according to the operations. - Abstract: This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. By carrying out tests and researches through a stress testing technique, P–σ curves in loading and unloading processes of the HDS are obtained and the stress coefficient k{sub f} of the HDS is obtained. So, the

  1. AUTOMATIC RECOGNITION OF CORONAL TYPE II RADIO BURSTS: THE AUTOMATED RADIO BURST IDENTIFICATION SYSTEM METHOD AND FIRST OBSERVATIONS

    International Nuclear Information System (INIS)

    Lobzin, Vasili V.; Cairns, Iver H.; Robinson, Peter A.; Steward, Graham; Patterson, Garth

    2010-01-01

    Major space weather events such as solar flares and coronal mass ejections are usually accompanied by solar radio bursts, which can potentially be used for real-time space weather forecasts. Type II radio bursts are produced near the local plasma frequency and its harmonic by fast electrons accelerated by a shock wave moving through the corona and solar wind with a typical speed of ∼1000 km s -1 . The coronal bursts have dynamic spectra with frequency gradually falling with time and durations of several minutes. This Letter presents a new method developed to detect type II coronal radio bursts automatically and describes its implementation in an extended Automated Radio Burst Identification System (ARBIS 2). Preliminary tests of the method with spectra obtained in 2002 show that the performance of the current implementation is quite high, ∼80%, while the probability of false positives is reasonably low, with one false positive per 100-200 hr for high solar activity and less than one false event per 10000 hr for low solar activity periods. The first automatically detected coronal type II radio burst is also presented.

  2. Development of ultrasonic testing technique with a large transducer to inspect the containment vessel plates embedded in concrete for corrosion on nuclear power plant (2)

    International Nuclear Information System (INIS)

    Ishida, Hitoshi

    2005-01-01

    The containment vessel plates embedded in concrete on Pressurized Water Reactors are inaccessible to inspect directly. Therefore, it is advisable to prepare inspection technology to detect existence and a location of corrosion on the embedded plates indirectly. The purpose of this study is establishment of ultrasonic testing technique to be able to inspect the containment vessel plates embedded in concrete widely from the accessible point. Experiments to detect artificial hollows simulating corrosion and stud bolts which hold the mold of concrete on a surface of a carbon steel plate mock-up covered with concrete were carried out with newly made low frequency (0.3MHz and 0.5MHz) 90 degrees refraction angle shear horizontal (SH) wave transducers combined with three active elements, which were equivalent to a 120 mm width element. As the results: (1) The echoes from the artificial hollows with a depth of 19 mm and 9.5mm at a distance of 1.5 m and the stud bolts with a diameter of 8mm at a distance of 0.7 - 1.7m could be discriminated clearly. (2) The multiple echoes bouncing three times between the front side and the back side of the plate, which was equivalent to a distance of about 12m, could be discriminated. (3) A divergence angle and a -6dB divergence angle of the large element (combined three elements) transducer were about 7 degrees and about 3 degrees. (4) The echoes from the hollows with a depth of 9.5m could be detected at a distance of 3.6 m with a reflection at the side wall of the mock-up. (5) It was estimated that the maximum distance of detection of the echo from the stud bolt with a diameter of 8mm was about 2.9 ∼ 3.6 m. Therefore we evaluate that the large element transducer can propagate the SH wave to about a half of a distance to the bottom of the embedded containment vessel and it is possible to detect the defects such as corrosion to a distance of 3.6 m. (author)

  3. Optical observations of Gamma-Ray Bursts

    International Nuclear Information System (INIS)

    Hjorth, J.; Pian, E.; Fynbo, J.P.U.

    2004-01-01

    We briefly review the status and recent progress in the field of optical observations of gamma-ray burst afterglows. We will focus on the fundamental observational evidence for the relationship between gamma-ray bursts and the final evolutionary phases of massive stars. In particular, we will address (i) gamma-ray burst host galaxies, (ii) optically dark gamma-ray burst afterglows, (iii) the gamma-ray burst-supernova connection, and (iv) the relation between X-ray flashes, gamma-ray bursts, and supernovae

  4. Enhancement of the quality of the reactor pressure vessel used in light water power plants by advanced material, fabrication and testing technologies

    International Nuclear Information System (INIS)

    Kussmaul, K.; Ewald, J.; Maier, G.; Schellhammer, W.

    1980-01-01

    Fracture safe assessment of nuclear reactor pressure vessels (RPV) is based upon an adequate stress analysis, reliable material characteristics, and acceptable defect sizes. Problems may arise concerning inhomogeneties, low toughness and crack phenomena as observed in the base material and heat affected zone (HAZ). Therefore, efforts have been made to develop a steel which would be both non-susceptible to embrittlement and/or cracking in the HAZ, and have a higher upper-shelf toughness of base and HAZ material. Tests have been made on inhomogeneties and defects and also on improvement of chemical composition, the steel-making process, welding procedures and the optimum temperature cycle and level for stress-relief heat treatment. To solve these problems, common testing methods were supplemented by tangential-cut techniques, small HAZ-tensile test procedures and HAZ-simulation techniques. Results indicate that 50 per cent of 100 investigated component-strength welds are affected by micro stress-relief cracking (SRC) on a micro-and millimetre scale. The 22 NiMoCr 37 steel with optimised chemical composition, and the 20 MnMoNi 55 steel are both resistant to stress-relief embrittlement and SRC. Specific welding techniques are found to limit SRC and proposals for optimum stress-relief temperatures are given. For the generation of new components, the fracture-safe analysis can now be based completely upon homogeneous and high upper-shelf base materials including the HAZ. (author)

  5. Vessel Operator System

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Operator cards are required for any operator of a charter/party boat and or a commercial vessel (including carrier and processor vessels) issued a vessel permit from...

  6. ASTM international symposium on small specimen test techniques and their applications to pressure vessel annealing and plant life extension

    International Nuclear Information System (INIS)

    Garner, F.A.; Hamilton, M.L.; Heinisch, H.L.; Kumar, A.S.

    1992-01-01

    Miniature sheet-type tensile specimens are currently being used in a variety of radiation damage studies conducted in a number of different reactors. Although these specimens are very small, they have proven successful in addressing issues encountered in both thermal reactors and anticipated fusion reactors. This paper reviews the results of a number of recent studies that illustrate the range of applicability of these small specimens. When combined with other types of specimens and other types of measurements made prior to tensile testing, miniature tensile specimens have been found to serve as very useful tools for application to both fundamental studies and alloy screening studies

  7. Cosmic gamma-ray bursts

    International Nuclear Information System (INIS)

    Hurley, K.

    1989-01-01

    This paper reviews the essential aspects of the gamma-ray burst (GRB) phenomenon, with emphasis on the more recent results. GRBs are introduced by their time histories, which provide some evidence for a compact object origin. The energy spectra of bursts are presented and they are seen to demonstrate practically unambiguously that the origin of some GRBs involves neutron stars. Counterpart searches are reviewed briefly and the statistical properties of bursters treated. This paper presents a review of the three known repeating bursters (the Soft Gamma Repeaters). Extragalactic and galactic models are discussed and future prospects are assessed

  8. Different Types of X-Ray Bursts from GRS 1915+105 and Their Origin

    Science.gov (United States)

    Yadav, J. S.; Rao, A. R.; Agrawal, P. C.; Paul, B.; Seetha, S.; Kasturirangan, K.

    1999-06-01

    We report X-ray observations of the Galactic X-ray transient source GRS 1915+105 with the pointed proportional counters of the Indian X-ray Astronomy Experiment (IXAE) onboard the Indian satellite IRS-P3, which show remarkable richness in temporal variability. The observations were carried out on 1997 June 12-29 and August 7-10, in the energy range of 2-18 keV and revealed the presence of very intense X-ray bursts. All the observed bursts have a slow exponential rise, a sharp linear decay, and broadly can be put in two classes: irregular and quasi-regular bursts in one class, and regular bursts in the other. The regular bursts are found to have two distinct timescales and to persist over extended durations. There is a strong correlation between the preceding quiescent time and the burst duration for the quasi-regular and irregular bursts. No such correlation is found for the regular bursts. The ratio of average flux during the burst time to the average flux during the quiescent phase is high and variable for the quasi-regular and irregular bursts, while it is low and constant for the regular bursts. We present a comprehensive picture of the various types of bursts observed in GRS 1915+105 in the light of the recent theories of advective accretion disks. We suggest that the peculiar bursts that we have seen are characteristic of the change of state of the source. The source can switch back and forth between the low-hard state and the high-soft state near critical accretion rates in a very short timescale, giving rise to the irregular and quasi-regular bursts. The fast timescale for the transition of the state is explained by invoking the appearance and disappearance of the advective disk in its viscous timescale. The periodicity of the regular bursts is explained by matching the viscous timescale with the cooling timescale of the postshock region. A test of the model is presented using the publicly available 13-60 keV RXTE/PCA data for irregular and regular bursts

  9. BATSE/OSSE Rapid Burst Response

    National Research Council Canada - National Science Library

    Matz, S. M; Grove, J. E; Johnson, W. N; Kurfess, J. D; Share, G. H; Fishman, G. J; Meegan, Charles A

    1995-01-01

    ...) slew the OSSE detectors to burst locations determined on-board by BATSE. This enables OSSE to make sensitive searches for prompt and delayed post-burst line and continuum emission above 50 keV...

  10. US Army Nuclear Burst Detection System (NBDS)

    International Nuclear Information System (INIS)

    Glaser, R.F.

    1980-07-01

    The Nuclear Burst Detection System (NBDS) was developed to meet the Army requirements of an unattended, automatic nuclear burst reporting system. It provides pertinent data for battlefield commanders on a timely basis with high reliability

  11. Short duration gamma ray bursts

    Indian Academy of Sciences (India)

    Observations have revealed that long bursts, with recorded afterglow, tend to reside in the star forming regions of normal galaxies. Moreover, GRB 980425 ... observer is negligible due to the special relativistic time dilation. However, because of deceleration, eventually Γ−1 > θj and thereafter, sideways expansion becomes.

  12. Optothermally actuated capillary burst valve

    DEFF Research Database (Denmark)

    Eriksen, Johan; Bilenberg, Brian; Kristensen, Anders

    2017-01-01

    be burst by raising the temperature due to the temperature dependence of the fluid surface tension. We address individual valves by using a local heating platform based on a thin film of near infrared absorber dye embedded in the lid used to seal the microfluidic device [L. H. Thamdrup et al., Nano Lett...

  13. Dark gamma-ray bursts

    Science.gov (United States)

    Brdar, Vedran; Kopp, Joachim; Liu, Jia

    2017-03-01

    Many theories of dark matter (DM) predict that DM particles can be captured by stars via scattering on ordinary matter. They subsequently condense into a DM core close to the center of the star and eventually annihilate. In this work, we trace DM capture and annihilation rates throughout the life of a massive star and show that this evolution culminates in an intense annihilation burst coincident with the death of the star in a core collapse supernova. The reason is that, along with the stellar interior, also its DM core heats up and contracts, so that the DM density increases rapidly during the final stages of stellar evolution. We argue that, counterintuitively, the annihilation burst is more intense if DM annihilation is a p -wave process than for s -wave annihilation because in the former case, more DM particles survive until the supernova. If among the DM annihilation products are particles like dark photons that can escape the exploding star and decay to standard model particles later, the annihilation burst results in a flash of gamma rays accompanying the supernova. For a galactic supernova, this "dark gamma-ray burst" may be observable in the Čerenkov Telescope Array.

  14. Stellar Sources of Gamma-ray Bursts

    OpenAIRE

    Luchkov, B. I.

    2011-01-01

    Correlation analysis of Swift gamma-ray burst coordinates and nearby star locations (catalog Gliese) reveals 4 coincidences with good angular accuracy. The random probability is 4\\times 10^{-5}, so evidencing that coincident stars are indeed gamma-ray burst sources. Some additional search of stellar gamma-ray bursts is discussed.

  15. Fine structure in fast drift storm bursts

    International Nuclear Information System (INIS)

    McConnell, D.; Ellis, G.R.A.

    1981-01-01

    Recent observations with high time resolution of fast drift storm (FDS) solar bursts are described. A new variety of FDS bursts characterised by intensity maxima regularly placed in the frequency domain is reported. Possible interpretations of this are mentioned and the implications of the short duration of FDS bursts are discussed. (orig.)

  16. Visual interface for the automation of the instrumented pendulum of Charpy tests used in the surveillance program of reactors vessel of nuclear power plants

    International Nuclear Information System (INIS)

    Rojas S, A.S.; Sainz M, E.; Ruiz E, J.A.

    2004-01-01

    Inside the Programs of Surveillance of the nuclear power stations periodic information is required on the state that keep the materials with those that builds the vessel of the reactor. This information is obtained through some samples or test tubes that are introduced inside the core of the reactor and it is observed if its physical characteristics remain after having been subjected to the radiation changes and temperature. The rehearsal with the instrumented Charpy pendulum offers information on the behavior of fracture dynamics of a material. In the National Institute of Nuclear Research (ININ) it has an instrumented Charpy pendulum. The operation of this instrument is manual, having inconveniences to carry out rehearsals with radioactive material, handling of high and low temperatures, to fulfill the normative ones for the realization of the rehearsals, etc. In this work the development of a computational program is presented (virtual instrument), for the automation of the instrumented pendulum. The system has modules like: Card of data acquisition, signal processing, positioning system, tempered system, pneumatic system, compute programs like it is the visual interface for the operation of the instrumented Charpy pendulum and the acquisition of impact signals. This system shows that given the characteristics of the nuclear industry with radioactive environments, the virtual instrumentation and the automation of processes can contribute to diminish the risks to the personnel occupationally exposed. (Author)

  17. Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2005-06-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-03, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. Total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and operator AM actions on steam generator (SG) secondary depressurization at a rate of -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes were assumed as experiment conditions. It is clarified that the AM actions are effective on primary system depressurization until the end of AIS injection at 1.6 MPa, but thereafter become less effective due to inflow of the non-condensable gas, resulting in delay of low pressure injection (LPI) actuation and whole core heatup under continuous water discharge through the bottom break. The report describes these thermohydraulic phenomena related with transient primary coolant mass and AM actions in addition to estimation of non-condensable gas behavior which affected primary-to-secondary heat transfer. (author)

  18. An experimental study on effective depressurization actions for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V test SB-PV-04)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2006-03-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westing-house-type PWR (3423 MWt). The experiment, SB-PV-04, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. It is clarified that AM actions with steam generator (SG) rapid depressurization by fully opening relief valves and auxiliary feedwater supply are effective to avoid core uncovery by actuating the low pressure injection (LPI) system though the primary depressurization is degraded by non-condensable gas inflow to the primary loops from the accumulator injection system. The effective core cooling was established by the rapid depressurization which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as AM actions. (author)

  19. Nuclear power plant prestressed concrete containment vessel structure monitoring during integrated leakage rate test using three kinds of fiber optic sensors

    Science.gov (United States)

    Liao, Kaixing; Li, Jinke; Kong, Xianglong; Sun, Changsen; Zhao, Xuefeng

    2017-04-01

    After years of operation, the safety of the prestressed concrete containment vessel (PCCV) structure of Nuclear Power Plant (NPP) is an important aspect. In order to detect the strength degradation and the structure deformation, several sensors such as vibrating wire strain gauge, invar wires and pendulums were installed in PCCV. However, the amounts of sensors above are limited due to the cost. Due to the well durability of fiber optic sensors, three kinds of fiber optic sensors were chosen to install on the surface of PCCV to monitor the deformation during Integrated Leakage Rate Test (ILRT). The three kinds of fiber optic sensors which had their own advantages and disadvantages are Fiber Bragg Grating (FBG), white light interferometry (WLI) and Brillouin Optical Time Domain Analysis (BOTDA). According to the measuring data, the three fiber optic sensors worked well during the ILRT. After the ILRT, the monitoring strain was recoverable thus the PCCV was still in the elastic stage. If these three kinds of fiber optic sensors are widely used in the PCCV, the unusual deformations are easier to detect. As a consequence, the three fiber optic sensors have good potential in the structure health monitoring of PCCV.

  20. Internally consistent gamma ray burst time history phenomenology

    International Nuclear Information System (INIS)

    Cline, T.L.

    1985-01-01

    A phenomenology for gamma ray burst time histories is outlined. Order of their generally chaotic appearance is attempted, based on the speculation that any one burst event can be represented above 150 keV as a superposition of similarly shaped increases of varying intensity. The increases can generally overlap, however, confusing the picture, but a given event must at least exhibit its own limiting characteristic rise and decay times if the measurements are made with instruments having adequate temporal resolution. Most catalogued observations may be of doubtful or marginal utility to test this hypothesis, but some time histories from Helios-2, Pioneer Venus Orbiter and other instruments having one-to several-millisecond capabilities appear to provide consistency. Also, recent studies of temporally resolved Solar Maximum Mission burst energy spectra are entirely compatible with this picture. The phenomenology suggested here, if correct, may assist as an analytic tool for modelling of burst processes and possibly in the definition of burst source populations

  1. Study on Monitoring Rock Burst through Drill Pipe Torque

    Directory of Open Access Journals (Sweden)

    Zhonghua Li

    2015-01-01

    Full Text Available This paper presents a new method to identify the danger of rock burst from the response of drill pipe torque during drilling process to overcome many defects of the conventional volume of drilled coal rubble method. It is based on the relationship of rock burst with coal stress and coal strength. Through theoretic analysis, the change mechanism of drill pipe torque and the relationship of drill pipe torque with coal stress, coal strength, and drilling speed are investigated. In light of the analysis, a new device for testing drill pipe torque is developed and a series of experiments is performed under different conditions; the results show that drill pipe torque linearly increases with the increase of coal stress and coal strength; the faster the drilling speed, the larger the drill pipe torque, and vice versa. When monitoring rock burst by drill pipe torque method, the index of rock burst is regarded as a function in which coal stress index and coal strength index are principal variables. The results are important for the forecast of rock burst in coal mine.

  2. Solar Radio Bursts and Space Weather

    Science.gov (United States)

    Gopalswamy, Natchimuthuk,

    2012-01-01

    Radio bursts from the Sun are produced by electron accelerated to relativistic energies by physical processes on the Sun such as solar flares and coronal mass ejections (CMEs). The radio bursts are thus good indicators of solar eruptions. Three types of nonthermal radio bursts are generally associated with CMEs. Type III bursts due to accelerated electrons propagating along open magnetic field lines. The electrons are thought to be accelerated at the reconnection region beneath the erupting CME, although there is another view that the electrons may be accelerated at the CME-driven shock. Type II bursts are due to electrons accelerated at the shock front. Type II bursts are also excellent indicators of solar energetic particle (SEP) events because the same shock is supposed accelerate electrons and ions. There is a hierarchical relationship between the wavelength range of type /I bursts and the CME kinetic energy. Finally, Type IV bursts are due to electrons trapped in moving or stationary structures. The low frequency stationary type IV bursts are observed occasionally in association with very fast CMEs. These bursts originate from flare loops behind the erupting CME and hence indicate tall loops. This paper presents a summary of radio bursts and their relation to CMEs and how they can be useful for space weather predictions.

  3. Dosimetry characterization of the Godiva Reactor under burst conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hickman, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Heinrichs, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hudson, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wong, C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ward, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilson, C. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Clark, L. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Trompier, F. [Inst. for Radiation Protection and Nuclear Safety, Fontenay-aux-Roses (France)

    2017-06-22

    A series of sixteen (16) burst irradiations were performed in May 2014, fifteen of which were part of an international collaboration to characterize the Godiva IV fast burst reactor at the National Criticality Experiments Research Center (NCERC). Godiva IV is a bare cylindrical assembly of approximately 65 kg of highly enriched uranium fuel (93.2% 235U metal alloyed with 1.5% molybdenum for strength) and is designed to perform controlled prompt critical excursions (Myers 2010, Goda 2013). Twelve of the irradiations were dedicated to neutron spectral measurements using a Bonner multiple sphere spectrometer. Three irradiations, with core temperature increases of 71.1°C, 136.9°C, and 229.9°C, were performed for generating comparative fluence data, establishing corrections for varying heights, testing linearity with burst temperature, and establishing gamma dose characteristics.

  4. Some polarization features of solar microwave bursts

    Energy Technology Data Exchange (ETDEWEB)

    Uralov, A M; Nefed' ev, V P [AN SSSR, Irkutsk. Sibirskij Inst. Zemnogo Magnetizma Ionosfery i Rasprostraneniya Radiovoln

    1977-01-01

    Consequences of the thermal microwave burst model proposed earlier have been considered. According to the model the centimeter burst is generated at the heat propagation to the upper atmosphere. The polarization features of the burst are explained: a change of the polarization sign in a frequency range, a rapid change of the polarization sign in the development of a burst at a fixed frequency, a lack of time coincidence of the moments of the burst maximum of the polarization and of the total flux. From the model the consequences are obtained, which are still not confirmed by experiment. An ordinary-type wave prevails in the burst radiation, in the course of which the polarization degree falls on the ascending branch of bursts development. At the change of the polarization sign at the fixed frequency prior to the sign change an ordinary-type wave should be present in excess and later an extreordinary type wave.

  5. Stimulus induced bursts in severe postanoxic encephalopathy.

    Science.gov (United States)

    Tjepkema-Cloostermans, Marleen C; Wijers, Elisabeth T; van Putten, Michel J A M

    2016-11-01

    To report on a distinct effect of auditory and sensory stimuli on the EEG in comatose patients with severe postanoxic encephalopathy. In two comatose patients admitted to the Intensive Care Unit (ICU) with severe postanoxic encephalopathy and burst-suppression EEG, we studied the effect of external stimuli (sound and touch) on the occurrence of bursts. In patient A bursts could be induced by either auditory or sensory stimuli. In patient B bursts could only be induced by touching different facial regions (forehead, nose and chin). When stimuli were presented with relatively long intervals, bursts persistently followed the stimuli, while stimuli with short intervals (encephalopathy can be induced by external stimuli, resulting in stimulus-dependent burst-suppression. Stimulus induced bursts should not be interpreted as prognostic favourable EEG reactivity. Copyright © 2016 International Federation of Clinical Neurophysiology. Published by Elsevier Ireland Ltd. All rights reserved.

  6. Predicting rock bursts in mines

    Science.gov (United States)

    Spall, H.

    1979-01-01

    In terms of lives lost, rock bursts in underground mines can be as hazardous as earthquakes on the surface. So it is not surprising that fo the last 40 years the U.S Bureau of Mines has been using seismic methods for detecting areas in underground mines where there is a high differential stress which could lead to structural instability of the rock mass being excavated.

  7. Coherent network analysis technique for discriminating gravitational-wave bursts from instrumental noise

    International Nuclear Information System (INIS)

    Chatterji, Shourov; Lazzarini, Albert; Stein, Leo; Sutton, Patrick J.; Searle, Antony; Tinto, Massimo

    2006-01-01

    The sensitivity of current searches for gravitational-wave bursts is limited by non-Gaussian, nonstationary noise transients which are common in real detectors. Existing techniques for detecting gravitational-wave bursts assume the output of the detector network to be the sum of a stationary Gaussian noise process and a gravitational-wave signal. These techniques often fail in the presence of noise nonstationarities by incorrectly identifying such transients as possible gravitational-wave bursts. Furthermore, consistency tests currently used to try to eliminate these noise transients are not applicable to general networks of detectors with different orientations and noise spectra. In order to address this problem we introduce a fully coherent consistency test that is robust against noise nonstationarities and allows one to distinguish between gravitational-wave bursts and noise transients in general detector networks. This technique does not require any a priori knowledge of the putative burst waveform

  8. NICER Eyes on Bursting Stars

    Science.gov (United States)

    Kohler, Susanna

    2018-03-01

    What happens to a neutron stars accretion disk when its surface briefly explodes? A new instrument recently deployed at the International Space Station (ISS) is now watching bursts from neutron stars and reporting back.Deploying a New X-Ray MissionLaunch of NICER aboard a Falcon 9 rocket in June 2017. [NASA/Tony Gray]In early June of 2017, a SpaceX Dragon capsule on a Falcon 9 rocket launched on a resupply mission to the ISS. The pressurized interior of the Dragon contained the usual manifest of crew supplies, spacewalk equipment, and vehicle hardware. But the unpressurized trunk of the capsule held something a little different: the Neutron star Interior Composition Explorer (NICER).In the two weeks following launch, NICER was extracted from the SpaceX Dragon capsule and installed on the ISS. And by the end of the month, the instrument was already collecting its first data set: observations of a bright X-ray burst from Aql X-1, a neutron star accreting matter from a low-mass binary companion.Impact of BurstsNICERs goal is to provide a new view of neutron-star physics at X-ray energies of 0.212 keV a window that allows us to explore bursts of energy that neutron stars sometimes emit from their surfaces.Artists impression of an X-ray binary, in which a compact object accretes material from a companion star. [ESA/NASA/Felix Mirabel]In X-ray burster systems, hydrogen- and helium-rich material from a low-mass companion star piles up in an accretion disk around the neutron star. This material slowly funnels onto the neutron stars surface, forming a layer that gravitationally compresses and eventually becomes so dense and hot that runaway nuclear fusion ignites.Within seconds, the layer of material is burned up, producing a burst of emission from the neutron star that outshines even the inner regions of the hot accretion disk. Then more material funnels onto the neutron star and the process begins again.Though we have a good picture of the physics that causes these bursts

  9. Bubble bursting at an interface

    Science.gov (United States)

    Kulkarni, Varun; Sajjad, Kumayl; Anand, Sushant; Fezzaa, Kamel

    2017-11-01

    Bubble bursting is crucial to understanding the life span of bubbles at an interface and more importantly the nature of interaction between the bulk liquid and the outside environment from the point of view of chemical and biological material transport. The dynamics of the bubble as it rises from inside the liquid bulk to its disappearance on the interface after bursting is an intriguing process, many aspects of which are still being explored. In our study, we make detailed high speed imaging measurements to examine carefully the hole initiation and growth in bursting bubbles that unearth some interesting features of the process. Previous analyses available in literature are revisited based on our novel experimental visualizations. Using a combination of experiments and theory we investigate the role of various forces during the rupturing process. This work aims to further our current knowledge of bubble dynamics at an interface with an aim of predicting better the bubble evolution from its growth to its eventual integration with the liquid bulk.

  10. Flaw distribution development from vessel ISI data

    International Nuclear Information System (INIS)

    Foulds, J.R.; Kennedy, E.L.; Basin, S.L.; Rosinski, S.T.

    1991-01-01

    Previous attempts to develop flaw distributions for use in the structural integrity evaluation of pressurized water reactor (PWR) vessels have aimed at the estimation of a ''generic'' distribution applicable to all vessels. In contrast, this paper describes the analysis of vessel-specific in-service inspection (ISI) data for the development of a flaw distribution reliably representative of the condition of the particular vessel inspected. The application of the methodology may be extended to other vessels, but has been primarily developed for PWR reactor vessels. For this study, the flaw data analyzed included data obtained from three recently performed PWR vessel ISIs and from laboratory inspection of selected weldment sections of the Midland reactor vessel. The variability in both the character of the reviewed data (size range of flaws, number of flaws) and the UT (ultrasonic test) inspection system performance identified a need for analyzing the inspection results on a vessel-, or data set-specific basis. For this purpose, traditional histogram-based methods were inadequate, and a new methodology that can accept a very small number of flaws (typical of vessel-specific ISI results) and that includes consideration of inspection system flaw detection reliability, flaw sizing accuracy and flaw detection threshold, was developed. Results of the application of the methodology to each of the four PWR reactor vessel cases studied are presented and discussed

  11. Study on closed pressure vessel test. Effect of heat rate, sample weight and vessel size on pressure rise due to thermal decomposition; Mippeigata atsuryoku yoki shiken ni kansuru kenkyu. Atsuryoku hassei kyodo ni oyobosu kanetsusokudo, shiryoryo oyobi youki saizu no eikyo

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Kenji.; Akutsu, Yoshiaki.; Arai, Mitsuru.; Tamura, Masamitsu. [The University of Tokyo, Tokyo (Japan). School of Engineering

    1999-02-28

    We have attempted to devise a new closed pressure vessel test apparatus in order to evaluate the violence of thermal decomposition of self-reactive materials and have examined some influencing factors, such as heat rate, sample weight, filling factor (sample weight/vessel size) and vessel size on Pmax (maximum pressure rise) and dP/dt (rate of pressure rise) due to their thermal decomposition. As a result, the following decreasing orders of Pmax and dP/dt were shown. Pmax: ADCA>BPZ>AIBN>TCP dP/dt: AIBN>BPZ>ADCA>TCP Moreover, Pmax was not almost influenced by heat rate, while dP/dt increased with an increase in heat rate in the case of BPZ. Pmax and dP/dt increased with an increase in sample weight and the degree of increase depended on the kinds of materials. In addition, it was shown that Pmax and dP/dt increased with an increase in vessel size at a constant filling factor. (author)

  12. Multiple shell pressure vessel

    International Nuclear Information System (INIS)

    Wedellsborg, B.W.

    1988-01-01

    A method is described of fabricating a pressure vessel comprising the steps of: attaching a first inner pressure vessel having means defining inlet and outlet openings to a top flange, placing a second inner pressure vessel, having means defining inlet and outlet opening, concentric with and spaced about the first inner pressure vessel and attaching the second inner pressure vessel to the top flange, placing an outer pressure vessel, having inlet and outlet openings, concentric with and spaced apart about the second inner pressure vessel and attaching the outer pressure vessel to the top flange, attaching a generally cylindrical inner inlet conduit and a generally cylindrical inner outlet conduit respectively to the inlet and outlet openings in the first inner pressure vessel, attaching a generally cylindrical outer inlet conduit and a generally cylindrical outer outlet conduit respectively to the inlet and outlet opening in the second inner pressure vessel, heating the assembled pressure vessel to a temperature above the melting point of a material selected from the group, lead, tin, antimony, bismuth, potassium, sodium, boron and mixtures thereof, filling the space between the first inner pressure vessel and the second inner pressure vessel with material selected from the group, filling the space between the second inner pressure vessel and the outer pressure vessel with material selected from the group, and pressurizing the material filling the spaces between the pressure vessels to a predetermined pressure, the step comprising: pressurizing the spaces to a pressure whereby the wall of the first inner pressure vessel is maintained in compression during steady state operation of the pressure vessel

  13. Neutron Assay System for Confinement Vessel Disposition

    International Nuclear Information System (INIS)

    Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Valdez, Jose I.; Vigil, Georgiana M.

    2012-01-01

    Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the CVs. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of special nuclear material (SNM) in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of (le)100-g 239 Pu equivalent in a vessel for safeguards termination. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements.

  14. Multi-rod burst behavior under a loss-of-coolant accident condition, (1)

    International Nuclear Information System (INIS)

    Hashimoto, Masao; Otomo, Takashi; Furuta, Teruo; Kawasaki, Satoru; Uetsuka, Hiroshi

    1980-12-01

    Multi-rod burst tests have been planned since 1977 to estimate quantitative channel restriction during a LOCA transient in LWRs. For this purpose, many bundle tests have been making to burst in a steam in varying a few parameters which influence the degree of channel restriction. The purpose of this report is to provide a background document for final reports to be published in the future. This report includes the results of No. 7805 bundle test, namely temperature, internal pressure, burst behavior of rods and channel restriction of the bundle. (author)

  15. Results of using engineering and technological measures for rock burst prevention. [USSR

    Energy Technology Data Exchange (ETDEWEB)

    Kulikov, A P; Nechaev, A V; Khmara, O I

    1980-01-01

    The paper evaluates methods for rock burst forecasting and rock burst prevention used in the Donbass, Kuzbass, Karaganda and Pechora basins. Forecasting methods are based on measuring the initial velocity of gas flow from test boreholes and/or quantity ratio of drillings leaving a test borehole and monitoring seismoacoustic signals. Number of working faces at which each of the methods for rock burst forecasting is used is given. Methods for rock burst prevention are comparatively evaluated: explosive fracturing of rocks in seam roof or seam floor, fluid injection (water and surfactants), drilling destressing boreholes, cutting destressing slots using cutting machines or water jets, mining protective coal seams first for reducing rock burst hazard in protected coal seams, using narrow web coal cutter loaders, remote control of coal cutters at working faces with extremely high rock burst hazard, using mining schemes which reduce rock burst hazards (e.g. long pillar mining system). From 1976 to 1979 number of rock bursts in underground coal mines in the USSR decreased by 5 times in comparison to the period 1961 to 1965. (3 refs.) (In Russian)

  16. Fuzzy-Based Adaptive Hybrid Burst Assembly Technique for Optical Burst Switched Networks

    Directory of Open Access Journals (Sweden)

    Abubakar Muhammad Umaru

    2014-01-01

    Full Text Available The optical burst switching (OBS paradigm is perceived as an intermediate switching technology for future all-optical networks. Burst assembly that is the first process in OBS is the focus of this paper. In this paper, an intelligent hybrid burst assembly algorithm that is based on fuzzy logic is proposed. The new algorithm is evaluated against the traditional hybrid burst assembly algorithm and the fuzzy adaptive threshold (FAT burst assembly algorithm via simulation. Simulation results show that the proposed algorithm outperforms the hybrid and the FAT algorithms in terms of burst end-to-end delay, packet end-to-end delay, and packet loss ratio.

  17. An investigation into the suitability of additive manufacturing techniques for neutron moderator vessels

    International Nuclear Information System (INIS)

    Gallimore, S.

    2016-01-01

    Additive manufacturing (also known as rapid prototyping or 3D printing) techniques are increasing in popularity for several key reasons; greater freedom in possible geometry, reduced time of manufacture and connected to these are potential cost savings. ISIS has begun an investigation into the suitability of the various available techniques for the manufacture of neutron moderator vessels, in order to see if it can exploit these advantages. It is however understood that additive manufacturing is by no means a perfect technique and part of the investigations will be to try and better understand how some of the disadvantages of the technique affect its potential application within the spallation neutron environment. Some of the main disadvantages commonly listed are; the grades of materials available/suitable for the process are limited, virtually no pre-existing material data from radiation environments, lower quality surface finish (directly from the manufacturing process), less familiarity with residual stresses in the material and questions over whether tight tolerances and consistent material thicknesses be achieved? The work has been divided into two streams; one which utilises small samples to evaluate and compare different manufacturing and post-treatment techniques, the other that performs tests on a full-size representative moderator vessel. The complete programme of testing shall include the following tests; fundamental 'neutronic transparency', room temperature vacuum leak test, cold shock (using LN_2) and subsequent room temperature leak test, pressure cycling, a burst test, welding suitability and material data testing. The investigations being conducted at ISIS are very much in the early stages and looking at fairly fundamental questions. Answering these will clearly guide the decision whether is it worth continuing with further investigation and development or if the currently available techniques do not produce materials that are suitable for use as

  18. Probabilistic retinal vessel segmentation

    Science.gov (United States)

    Wu, Chang-Hua; Agam, Gady

    2007-03-01

    Optic fundus assessment is widely used for diagnosing vascular and non-vascular pathology. Inspection of the retinal vasculature may reveal hypertension, diabetes, arteriosclerosis, cardiovascular disease and stroke. Due to various imaging conditions retinal images may be degraded. Consequently, the enhancement of such images and vessels in them is an important task with direct clinical applications. We propose a novel technique for vessel enhancement in retinal images that is capable of enhancing vessel junctions in addition to linear vessel segments. This is an extension of vessel filters we have previously developed for vessel enhancement in thoracic CT scans. The proposed approach is based on probabilistic models which can discern vessels and junctions. Evaluation shows the proposed filter is better than several known techniques and is comparable to the state of the art when evaluated on a standard dataset. A ridge-based vessel tracking process is applied on the enhanced image to demonstrate the effectiveness of the enhancement filter.

  19. Improvement to reactor vessel

    International Nuclear Information System (INIS)

    1974-01-01

    The vessel described includes a prestressed concrete vessel containing a chamber and a removable cover closing this chamber. The cover is in concrete and is kept in its closed position by main and auxiliary retainers, comprising fittings integral with the concrete of the vessel. The auxiliary retainers pass through the concrete of the cover. This improvement may be applied to BWR, PWR and LMFBR type reactor vessel [fr

  20. Transitions to Synchrony in Coupled Bursting Neurons

    Science.gov (United States)

    Dhamala, Mukeshwar; Jirsa, Viktor K.; Ding, Mingzhou

    2004-01-01

    Certain cells in the brain, for example, thalamic neurons during sleep, show spike-burst activity. We study such spike-burst neural activity and the transitions to a synchronized state using a model of coupled bursting neurons. In an electrically coupled network, we show that the increase of coupling strength increases incoherence first and then induces two different transitions to synchronized states, one associated with bursts and the other with spikes. These sequential transitions to synchronized states are determined by the zero crossings of the maximum transverse Lyapunov exponents. These results suggest that synchronization of spike-burst activity is a multi-time-scale phenomenon and burst synchrony is a precursor to spike synchrony.

  1. Transitions to synchrony in coupled bursting neurons

    International Nuclear Information System (INIS)

    Dhamala, Mukeshwar; Jirsa, Viktor K.; Ding Mingzhou

    2004-01-01

    Certain cells in the brain, for example, thalamic neurons during sleep, show spike-burst activity. We study such spike-burst neural activity and the transitions to a synchronized state using a model of coupled bursting neurons. In an electrically coupled network, we show that the increase of coupling strength increases incoherence first and then induces two different transitions to synchronized states, one associated with bursts and the other with spikes. These sequential transitions to synchronized states are determined by the zero crossings of the maximum transverse Lyapunov exponents. These results suggest that synchronization of spike-burst activity is a multi-time-scale phenomenon and burst synchrony is a precursor to spike synchrony

  2. Swift Burst Alert Telescope (BAT) Instrument Response

    International Nuclear Information System (INIS)

    Parsons, A.; Barthelmy, S.; Cummings, J.; Gehrels, N.; Hullinger, D.; Krimm, H.; Markwardt, C.; Tueller, J.; Fenimore, E.; Palmer, D.; Sato, G.; Takahashi, T.; Nakazawa, K.; Okada, Y.; Takahashi, H.; Suzuki, M.; Tashiro, M.

    2004-01-01

    The Burst Alert Telescope (BAT), a large coded aperture instrument with a wide field-of-view (FOV), provides the gamma-ray burst triggers and locations for the Swift Gamma-Ray Burst Explorer. In addition to providing this imaging information, BAT will perform a 15 keV - 150 keV all-sky hard x-ray survey based on the serendipitous pointings resulting from the study of gamma-ray bursts, and will also monitor the sky for transient hard x-ray sources. For BAT to provide spectral and photometric information for the gamma-ray bursts, the transient sources and the all-sky survey, the BAT instrument response must be determined to an increasingly greater accuracy. This paper describes the spectral models and the ground calibration experiments used to determine the BAT response to an accuracy suitable for gamma-ray burst studies

  3. ALICE HMPID Radiator Vessel

    CERN Document Server

    2003-01-01

    View of the radiator vessels of the ALICE/HMPID mounted on the support frame. Each HMPID module is equipped with 3 indipendent radiator vessels made out of neoceram and fused silica (quartz) windows glued together. The spacers inside the vessel are needed to stand the hydrostatic pressure. http://alice-hmpid.web.cern.ch/alice-hmpid

  4. Bursting in the Belousov-Zhabotinsky Reaction added with Phenol in a Batch Reactor

    International Nuclear Information System (INIS)

    Cadena, Ariel; Agreda, Jesus; Barragan, Daniel

    2013-01-01

    The classic Belousov-Zhabotinski reaction was modified by adding phenol as a second organic substrate that kinetically competes with the malonic acid in the reduction of Ce 4+ to Ce 3+ and in the removal of molecular bromine of the reaction mixture. The oscillating reaction of two substrates exhibited burst firing and an oscillatory period of long duration. Analysis of experimental data shows an increasing of the bursting phenomenon, with a greater spiking in the burst firing and with a longer quiescent state, as a function of the initial phenol concentration increase. It was hypothesized that the bursting phenomenon can be explained introducing a redox cycle between the reduced phenolic species (hydroxyphenols) and the oxidized ones (quinones). The hypothesis was experimentally and numerically tested and from the results it is possible to conclude that the bursting phenomenon exhibited by the oscillating reaction of two substrates is mainly driven by a p-di-hydroxy-benzene/p-benzoquinone redox cycle (author)

  5. ROSA-V/LSTF vessel top head LOCA tests SB-PV-07 and SB-PV-08 with break sizes of 1.0 and 0.1% and operator recovery actions for core cooling

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    2010-02-01

    A series of break size parameter tests (SB-PV-07 and SB-PV-08) were conducted at the Large Scale Test Facility (LSTF) of ROSA-V Program by simulating a vessel top small break loss-of-coolant accident (SBLOCA) at a pressurized water reactor (PWR). Typical phenomena to the vessel top break LOCA and effectiveness of operator recovery actions on core cooling were studied under an assumption of total failure of high pressure injection (HPI) system. The LSTF simulates a 4-loop 3423 MWt PWR by a full-height, full-pressure and 1/48 volume scaling two-loop system. Typical phenomena of vessel top break LOCA are clarified for the cases with break sizes of 1.0 and 0.1% cold leg break equivalent. The results from a 0.5% top break LOCA test (SB-PV-02) in the early ROSA-IV Program was referred during discussion. Operator actions of HPI recovery in the 1.0% top break test and steam generator (SG) depressurization in the 0.1% top break test were initiated when temperature at core exit thermocouple (CET) reached 623 K during core boil-off. Both operator actions resulted in immediate recovery of core cooling. Based on the obtained data, several thermal-hydraulic phenomena were discussed further such as relations between vessel top head water level and steam discharge at the break, and between coolant mass inventory transient and core heat-up and quench behavior, and CET performances to detect core heat-up under influences of three-dimensional (3D) steam flows in the core and core exit. (author)

  6. Detection of artifacts from high energy bursts in neonatal EEG.

    Science.gov (United States)

    Bhattacharyya, Sourya; Biswas, Arunava; Mukherjee, Jayanta; Majumdar, Arun Kumar; Majumdar, Bandana; Mukherjee, Suchandra; Singh, Arun Kumar

    2013-11-01

    Detection of non-cerebral activities or artifacts, intermixed within the background EEG, is essential to discard them from subsequent pattern analysis. The problem is much harder in neonatal EEG, where the background EEG contains spikes, waves, and rapid fluctuations in amplitude and frequency. Existing artifact detection methods are mostly limited to detect only a subset of artifacts such as ocular, muscle or power line artifacts. Few methods integrate different modules, each for detection of one specific category of artifact. Furthermore, most of the reference approaches are implemented and tested on adult EEG recordings. Direct application of those methods on neonatal EEG causes performance deterioration, due to greater pattern variation and inherent complexity. A method for detection of a wide range of artifact categories in neonatal EEG is thus required. At the same time, the method should be specific enough to preserve the background EEG information. The current study describes a feature based classification approach to detect both repetitive (generated from ECG, EMG, pulse, respiration, etc.) and transient (generated from eye blinking, eye movement, patient movement, etc.) artifacts. It focuses on artifact detection within high energy burst patterns, instead of detecting artifacts within the complete background EEG with wide pattern variation. The objective is to find true burst patterns, which can later be used to identify the Burst-Suppression (BS) pattern, which is commonly observed during newborn seizure. Such selective artifact detection is proven to be more sensitive to artifacts and specific to bursts, compared to the existing artifact detection approaches applied on the complete background EEG. Several time domain, frequency domain, statistical features, and features generated by wavelet decomposition are analyzed to model the proposed bi-classification between burst and artifact segments. A feature selection method is also applied to select the

  7. Chaotic bursting in semiconductor lasers

    Science.gov (United States)

    Ruschel, Stefan; Yanchuk, Serhiy

    2017-11-01

    We investigate the dynamic mechanisms for low frequency fluctuations in semiconductor lasers subjected to delayed optical feedback, using the Lang-Kobayashi model. This system of delay differential equations displays pronounced envelope dynamics, ranging from erratic, so called low frequency fluctuations to regular pulse packages, if the time scales of fast oscillations and envelope dynamics are well separated. We investigate the parameter regions where low frequency fluctuations occur and compute their Lyapunov spectra. Using the geometric singular perturbation theory, we study this intermittent chaotic behavior and characterize these solutions as bursting slow-fast oscillations.

  8. Helium bubble bursting in tungsten

    International Nuclear Information System (INIS)

    Sefta, Faiza; Juslin, Niklas; Wirth, Brian D.

    2013-01-01

    Molecular dynamics simulations have been used to systematically study the pressure evolution and bursting behavior of sub-surface helium bubbles and the resulting tungsten surface morphology. This study specifically investigates how bubble shape and size, temperature, tungsten surface orientation, and ligament thickness above the bubble influence bubble stability and surface evolution. The tungsten surface is roughened by a combination of adatom “islands,” craters, and pinholes. The present study provides insight into the mechanisms and conditions leading to various tungsten topology changes, which we believe are the initial stages of surface evolution leading to the formation of nanoscale fuzz

  9. Compliant electrospun silk fibroin tubes for small vessel bypass grafting.

    Science.gov (United States)

    Marelli, Benedetto; Alessandrino, Antonio; Farè, Silvia; Freddi, Giuliano; Mantovani, Diego; Tanzi, Maria Cristina

    2010-10-01

    Processing silk fibroin (SF) by electrospinning offers a very attractive opportunity for producing three-dimensional nanofibrillar matrices in tubular form, which may be useful for a biomimetic approach to small calibre vessel regeneration. Bypass grafting of small calibre vessels, with a diameter less than 6mm, is performed mainly using autografts, like the saphenous vein or internal mammary artery. At present no polymeric grafts made of SF are commercially available, mainly due to inadequate properties (low compliance and lack of endothelium cells). The aim of this work was to electrospin SF into tubular structures (Ø=6mm) for small calibre vessel grafting, characterize the morphological, chemico-physical and mechanical properties of the electrospun SF structures and to validate their potential to interact with cells. The morphological properties of electrospun SF nanofibres were investigated by scanning electron microscopy. Chemico-physical analyses revealed an increase in the crystallinity of the structure of SF nanofibres on methanol treatment. Mechanical tests, i.e. compliance and burst pressure measurements, of the electrospun SF tubes showed that the inner pressure to radial deformation ratio was linear for elongation up to 15% and pressure up to 400 mm Hg. The mean compliance value between 80 and 120 mm Hg was higher than the values reported for both Goretex(R) and Dacron(R) grafts and for bovine heterografts, but still slightly lower than those of saphenous and umbilical vein, which nowadays represent the gold standard for the replacement of small calibre arteries. The electrospun tubes resisted up to 575+/-17 mmHg, which is more than four times the upper physiological pressure of 120 mmHg and more than twice the pathological upper pressures (range 180-220 mmHg). The in vitro tests showed a good cytocompatibility of the electrospun SF tubes. Therefore, the electrospun SF tubes developed within this work represent a suitable candidate for small calibre

  10. Selected elements of rock burst state assessment in case studies from the Silesian hard coal mines

    Energy Technology Data Exchange (ETDEWEB)

    Jozef Kabiesz; Janusz Makowka [Central Mining Institute, Katowice (Poland)

    2009-09-15

    Exploitation of coal seams in the Upper Silesian Coal Basin is conducted in complex and difficult conditions. These difficulties are connected with the occurrence of many natural mining hazards and limitations resulting from the existing in this area surface infrastructure. One of the most important problems of Polish mining is the rock burst hazard and reliable evaluation of its condition. During long-years' mining practice in Poland a comprehensive system of evaluation and control of this hazard was developed. In the paper the main aspects of rock burst hazard state evaluation will be presented, comprising: 1) rock mass inclination for rock bursts, i.e., rock strength properties investigation, comprehensive parametric evaluation of rock mass inclination for rock bursts, prognosis of seismic events induced by mining operations, methods of computer-aided modelling of stress and rock mass deformation parameters distribution, strategic rock mass classification under rock burst degrees; 2) immediate seismic and rock burst hazard state evaluation, i.e., low diameter test drilling method, seismologic and seismoacoustic method, comprehensive method of rock burst hazard state evaluation, non-standard methods of evaluation; 3) legal aspects of rock burst hazard state evaluation. Selected elements of the hazard state evaluation system are illustrated with specific practical examples of their application. 11 refs., 14 figs.

  11. Neutrino burst identification in underground detectors

    International Nuclear Information System (INIS)

    Fulgione, W.; Mengotti-Silva, N.; Panaro, L.

    1996-01-01

    We discuss the problem of neutrino burst identification in underground ν-telescopes. First the usual statistical analysis based on the time structure of the events is reviewed, with special attention to the statistical significance of burst candidates. Next, we propose a second level analysis that can provide independent confirmation of burst detection. This exploits the spatial distribution of the single events of a burst candidate, and uses the formalism of the entropy of information. Examples of both techniques are shown, based on the LVD experiment at Gran Sasso. (orig.)

  12. Hydrogen storage in insulated pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S.M.; Garcia-Villazana, O. [Lawrence Livermore National Lab., CA (United States)

    1998-08-01

    Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LH{sub 2}) or ambient-temperature compressed hydrogen (CH{sub 2}). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (lower energy requirement for hydrogen liquefaction and reduced evaporative losses). This paper shows an evaluation of the applicability of the insulated pressure vessels for light-duty vehicles. The paper shows an evaluation of evaporative losses and insulation requirements and a description of the current analysis and experimental plans for testing insulated pressure vessels. The results show significant advantages to the use of insulated pressure vessels for light-duty vehicles.

  13. ASME Section VIII Recertification of a 33,000 Gallon Vacuum-jacketed LH2 Storage Vessel for Densified Hydrogen Testing at NASA Kennedy Space Center

    Science.gov (United States)

    Swanger, Adam M.; Notardonato, William U.; Jumper, Kevin M.

    2015-01-01

    The Ground Operations Demonstration Unit for Liquid Hydrogen (GODU-LH2) has been developed at NASA Kennedy Space Center in Florida. GODU-LH2 has three main objectives: zero-loss storage and transfer, liquefaction, and densification of liquid hydrogen. A cryogenic refrigerator has been integrated into an existing, previously certified, 33,000 gallon vacuum-jacketed storage vessel built by Minnesota Valley Engineering in 1991 for the Titan program. The dewar has an inner diameter of 9.5 and a length of 71.5; original design temperature and pressure ranges are -423 F to 100 F and 0 to 95 psig respectively. During densification operations the liquid temperature will be decreased below the normal boiling point by the refrigerator, and consequently the pressure inside the inner vessel will be sub-atmospheric. These new operational conditions rendered the original certification invalid, so an effort was undertaken to recertify the tank to the new pressure and temperature requirements (-12.7 to 95 psig and -433 F to 100 F respectively) per ASME Boiler and Pressure Vessel Code, Section VIII, Division 1. This paper will discuss the unique design, analysis and implementation issues encountered during the vessel recertification process.

  14. 46 CFR 115.812 - Pressure vessels and boilers.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Pressure vessels and boilers. 115.812 Section 115.812... CERTIFICATION Material Inspections § 115.812 Pressure vessels and boilers. (a) Pressure vessels must be tested... testing requirements for boilers are contained in § 61.05 in subchapter F of this chapter. [CGD 85-080, 61...

  15. Elastic plastic buckling of elliptical vessel heads

    International Nuclear Information System (INIS)

    Alix, M.; Roche, R.L.

    1981-08-01

    The risks of buckling of dished vessel head increase when the vessel is thin walled. This paper gives the last results on experimental tests of 3 elliptical heads and compares all the results with some empirical formula dealing with elastic and plastic buckling

  16. Development of Catamaran Fishing Vessel

    Directory of Open Access Journals (Sweden)

    A. Jamaluddin

    2010-11-01

    Full Text Available Multihull due to a couple of advantages has been the topic of extensive research work in naval architecture. In this study, a series of investigation of fishing vessel to save fuel energy was carried out at ITS. Two types of ship models, monohull (round bilge and hard chine and catamaran, a boat with two hulls (symmetrical and asymmetrical were developed. Four models were produced physically and numerically, tested (towing tank and simulated numerically (CFD code. The results of the two approaches indicated that the catamaran mode might have drag (resistance smaller than those of monohull at the same displacement. A layout of catamaran fishing vessel, proposed here, indicates the freedom of setting the deck equipments for fishing vessel.

  17. Analysis of historic bursts and burst detection in water supply areas of different size

    NARCIS (Netherlands)

    Bakker, M.; Trietsch, E.A.; Vreeburg, J.H.G.; Rietveld, L.C.

    2014-01-01

    Pipe bursts in water distribution networks lead to water losses and a risk of damaging the urban environment. We studied hydraulic data and customer contact records of 44 real bursts for a better understanding of the phenomena. We found that most bursts were reported to the water company shortly

  18. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  19. Supporting differentiated quality of service in optical burst switched networks

    Science.gov (United States)

    Zhou, Bin; Bassiouni, Mostafa A.

    2006-01-01

    We propose and evaluate two new schemes for providing differentiated services in optical burst switched (OBS) networks. The two new schemes are suitable for implementation in OBS networks using just-in-time (JIT) or just-enough-time (JET) scheduling protocols. The first scheme adjusts the size of the search space for a free wavelength based on the priority level of the burst. A simple equation is used to divide the search spectrum into two parts: a base part and an adjustable part. The size of the adjustable part increases as the priority of the burst becomes higher. The scheme is very easy to implement and does not demand any major software or hardware resources in optical cross-connects. The second scheme reduces the dropping probability of bursts with higher priorities through the use of different proactive discarding rates in the network access station (NAS) of the source node. Our extensive simulation tests using JIT show that both schemes are capable of providing tangible quality of service (QoS) differentiation without negatively impacting the throughput of OBS networks.

  20. Bursting as a source of non-linear determinism in the firing patterns of nigral dopamine neurons.

    Science.gov (United States)

    Jeong, Jaeseung; Shi, Wei-Xing; Hoffman, Ralph; Oh, Jihoon; Gore, John C; Bunney, Benjamin S; Peterson, Bradley S

    2012-11-01

    Nigral dopamine (DA) neurons in vivo exhibit complex firing patterns consisting of tonic single-spikes and phasic bursts that encode information for certain types of reward-related learning and behavior. Non-linear dynamical analysis has previously demonstrated the presence of a non-linear deterministic structure in complex firing patterns of DA neurons, yet the origin of this non-linear determinism remains unknown. In this study, we hypothesized that bursting activity is the primary source of non-linear determinism in the firing patterns of DA neurons. To test this hypothesis, we investigated the dimension complexity of inter-spike interval data recorded in vivo from bursting and non-bursting DA neurons in the chloral hydrate-anesthetized rat substantia nigra. We found that bursting DA neurons exhibited non-linear determinism in their firing patterns, whereas non-bursting DA neurons showed truly stochastic firing patterns. Determinism was also detected in the isolated burst and inter-burst interval data extracted from firing patterns of bursting neurons. Moreover, less bursting DA neurons in halothane-anesthetized rats exhibited higher dimensional spiking dynamics than do more bursting DA neurons in chloral hydrate-anesthetized rats. These results strongly indicate that bursting activity is the main source of low-dimensional, non-linear determinism in the firing patterns of DA neurons. This finding furthermore suggests that bursts are the likely carriers of meaningful information in the firing activities of DA neurons. © 2012 The Authors. European Journal of Neuroscience © 2012 Federation of European Neuroscience Societies and Blackwell Publishing Ltd.

  1. Structural Properties of EB-Welded AlSi10Mg Thin-Walled Pressure Vessels Produced by AM-SLM Technology

    Science.gov (United States)

    Nahmany, Moshe; Stern, Adin; Aghion, Eli; Frage, Nachum

    2017-10-01

    Additive manufacturing of metals by selective laser melting (AM-SLM) is hampered by significant limitations in product size due to the limited dimensions of printing trays. Electron beam welding (EBW) is a well-established process that results in relatively minor metallurgical modifications in workpieces due to the ability of EBW to pass high-density energy to the related substance. The present study aims to evaluate structural properties of EB-welded AlSi10Mg thin-walled pressure vessels produced from components prepared by SLM technology. Following the EB welding process, leak and burst tests were conducted, as was fractography analysis. The welded vessels showed an acceptable holding pressure of 30 MPa, with a reasonable residual deformation up to 2.3% and a leak rate better than 1 × 10-8 std-cc s-1 helium. The failures that occurred under longitudinal stresses reflected the presence of two weak locations in the vessels, i.e., the welded joint region and the transition zone between the vessel base and wall. Fractographic analysis of the fracture surfaces of broken vessels displayed the ductile mode of the rupture, with dimples of various sizes, depending on the failure location.

  2. Multifrequency Observations of Gamma-Ray Burst

    OpenAIRE

    Greiner, J.

    1995-01-01

    Neither a flaring nor a quiescent counterpart to a gamma-ray burst has yet been convincingly identified at any wavelength region. The present status of the search for counterparts of classical gamma-ray bursts is given. Particular emphasis is put on the search for flaring counterparts, i.e. emission during or shortly after the gamma-ray emission.

  3. Observations of short gamma-ray bursts.

    Science.gov (United States)

    Fox, Derek B; Roming, Peter W A

    2007-05-15

    We review recent observations of short-hard gamma-ray bursts and their afterglows. The launch and successful ongoing operations of the Swift satellite, along with several localizations from the High-Energy Transient Explorer mission, have provoked a revolution in short-burst studies: first, by quickly providing high-quality positions to observers; and second, via rapid and sustained observations from the Swift satellite itself. We make a complete accounting of Swift-era short-burst localizations and proposed host galaxies, and discuss the implications of these observations for the distances, energetics and environments of short bursts, and the nature of their progenitors. We then review the physical modelling of short-burst afterglows: while the simplest afterglow models are inadequate to explain the observations, there have been several notable successes. Finally, we address the case of an unusual burst that threatens to upset the simple picture in which long bursts are due to the deaths of massive stars, and short bursts to compact-object merger events.

  4. Neutrino bursts and gravitational waves experiments

    Energy Technology Data Exchange (ETDEWEB)

    Castagnoli, C; Galeotti, P; Saavedra, O [Consiglio Nazionale delle Ricerche, Turin (Italy). Lab. di Cosmo-Geofisica

    1978-05-01

    Several experiments have been performed in many countries to observe gravitational waves or neutrino bursts. Since their simultaneous emission may occur in stellar collapse, the authors evaluate the effect of neutrino bursts on gravitational wave antennas and suggest the usefulness of a time correlation among the different detectors.

  5. Polarization of a periodic solar microwave burst

    Energy Technology Data Exchange (ETDEWEB)

    Kaufmann, P [Universidade Mackenzie, Sao Paulo (Brazil). Centro de Radio-Astronomia e Astrofisica

    1976-09-01

    No fluctuations in polarization have been found during a 7 GHz solar burst showing 17s periodic pulses in intensity. Polarization effects can be produced by the propagation media in the active centre, which are not affected directly by the burst source, but situated more deeply than the observed heights at that microwave frequency.

  6. Estimates of the burst reliability of thin-walled cylinders designed to meet the ASME Code allowables

    International Nuclear Information System (INIS)

    Stancampiano, P.A.; Zemanick, P.P.

    1976-01-01

    Pressure containment components in nuclear power plants are designed by the conventional deterministic safety factor approach to meet the requirements of the ASME Pressure Vessel Code, Section III. The inevitable variabilities and uncertainties associated with the design, manufacture, installation, and service processes suggest a probabilistic design approach may also be pertinent. Accordingly, the burst reliabilities of two thin-walled 304 SS cylindrical vessels such as might be employed in liquid metal plants are estimated. A large vessel fabricated from rolled plate per ASME SA-240 and a smaller pipe sized vessel also fabricated from rolled plate per ASME SA-358 are considered. The vessels are sized to just meet the allowable ASME Code primary membrance stresses at 800 0 F (427 0 C). The bursting probability that the operating pressure is greater than the burst strength of the cylinders is calculated using stress-strength interference theory by direct Monte Carlo simulation on a high speed digital computer. A sensitivity study is employed to identify those design parameters which have the greatest effect on the reliability. The effects of preservice quality assurance defect inspections on the reliability are also evaluated parametrically

  7. Burst suppression probability algorithms: state-space methods for tracking EEG burst suppression

    Science.gov (United States)

    Chemali, Jessica; Ching, ShiNung; Purdon, Patrick L.; Solt, Ken; Brown, Emery N.

    2013-10-01

    Objective. Burst suppression is an electroencephalogram pattern in which bursts of electrical activity alternate with an isoelectric state. This pattern is commonly seen in states of severely reduced brain activity such as profound general anesthesia, anoxic brain injuries, hypothermia and certain developmental disorders. Devising accurate, reliable ways to quantify burst suppression is an important clinical and research problem. Although thresholding and segmentation algorithms readily identify burst suppression periods, analysis algorithms require long intervals of data to characterize burst suppression at a given time and provide no framework for statistical inference. Approach. We introduce the concept of the burst suppression probability (BSP) to define the brain's instantaneous propensity of being in the suppressed state. To conduct dynamic analyses of burst suppression we propose a state-space model in which the observation process is a binomial model and the state equation is a Gaussian random walk. We estimate the model using an approximate expectation maximization algorithm and illustrate its application in the analysis of rodent burst suppression recordings under general anesthesia and a patient during induction of controlled hypothermia. Main result. The BSP algorithms track burst suppression on a second-to-second time scale, and make possible formal statistical comparisons of burst suppression at different times. Significance. The state-space approach suggests a principled and informative way to analyze burst suppression that can be used to monitor, and eventually to control, the brain states of patients in the operating room and in the intensive care unit.

  8. Bursting synchronization in scale-free networks

    International Nuclear Information System (INIS)

    Batista, C.A.S.; Batista, A.M.; Pontes, J.C.A. de; Lopes, S.R.; Viana, R.L.

    2009-01-01

    Neuronal networks in some areas of the brain cortex present the scale-free property, i.e., the neuron connectivity is distributed according to a power-law, such that neurons are more likely to couple with other already well-connected ones. Neuron activity presents two timescales, a fast one related to action-potential spiking, and a slow timescale in which bursting takes place. Some pathological conditions are related with the synchronization of the bursting activity in a weak sense, meaning the adjustment of the bursting phase due to coupling. Hence it has been proposed that an externally applied time-periodic signal be applied in order to control undesirable synchronized bursting rhythms. We investigated this kind of intervention using a two-dimensional map to describe neurons with spiking-bursting activity in a scale-free network.

  9. X-Ray Bursts from NGC 6652

    Science.gov (United States)

    Morgan, Edward

    The possibly transient X-ray Source in the globular cluster NGC 6652 has been seen by BeppoSax and the ASM on RXTE to undergo X-ray bursts, possibly Type I. Very little is known about this X-ray source, and confirmation of its bursts type-I nature would identify it as a neutron star binary. Type I bursts in 6 other sources have been shown to exhibit intervals of millisecond ocsillation that most likely indicate the neutron star spin period. Radius-expansion bursts can reveal information about the mass and size of the neutron star. We propose to use the ASM to trigger an observation of this source to maximize the probability of catching a burst in the PCA.

  10. Pressure vessel design manual

    CERN Document Server

    Moss, Dennis R

    2013-01-01

    Pressure vessels are closed containers designed to hold gases or liquids at a pressure substantially different from the ambient pressure. They have a variety of applications in industry, including in oil refineries, nuclear reactors, vehicle airbrake reservoirs, and more. The pressure differential with such vessels is dangerous, and due to the risk of accident and fatality around their use, the design, manufacture, operation and inspection of pressure vessels is regulated by engineering authorities and guided by legal codes and standards. Pressure Vessel Design Manual is a solutions-focused guide to the many problems and technical challenges involved in the design of pressure vessels to match stringent standards and codes. It brings together otherwise scattered information and explanations into one easy-to-use resource to minimize research and take readers from problem to solution in the most direct manner possible. * Covers almost all problems that a working pressure vessel designer can expect to face, with ...

  11. Spatiotemporal chaos from bursting dynamics

    International Nuclear Information System (INIS)

    Berenstein, Igal; De Decker, Yannick

    2015-01-01

    In this paper, we study the emergence of spatiotemporal chaos from mixed-mode oscillations, by using an extended Oregonator model. We show that bursting dynamics consisting of fast/slow mixed mode oscillations along a single attractor can lead to spatiotemporal chaotic dynamics, although the spatially homogeneous solution is itself non-chaotic. This behavior is observed far from the Hopf bifurcation and takes the form of a spatiotemporal intermittency where the system locally alternates between the fast and the slow phases of the mixed mode oscillations. We expect this form of spatiotemporal chaos to be generic for models in which one or several slow variables are coupled to activator-inhibitor type of oscillators

  12. High energy neutrinos from gamma-ray bursts with precursor supernovae.

    Science.gov (United States)

    Razzaque, Soebur; Mészáros, Peter; Waxman, Eli

    2003-06-20

    The high energy neutrino signature from proton-proton and photo-meson interactions in a supernova remnant shell ejected prior to a gamma-ray burst provides a test for the precursor supernova, or supranova, model of gamma-ray bursts. Protons in the supernova remnant shell and photons entrapped from a supernova explosion or a pulsar wind from a fast-rotating neutron star remnant provide ample targets for protons escaping the internal shocks of the gamma-ray burst to interact and produce high energy neutrinos. We calculate the expected neutrino fluxes, which can be detected by current and future experiments.

  13. Quantum-Gravity Based Photon Dispersion in Gamma-Ray Bursts: The Detection Problem

    International Nuclear Information System (INIS)

    Norris, Jay P.; Scargle, Jeffrey D.

    2007-01-01

    Gamma-ray bursts at cosmological distances offer a time-varying signal that can be used to search for energy-dependent photon dispersion effects. We show that short bursts with narrow pulse structures at high energies will offer the least ambiguous tests for energy-dependent dispersion effects. We discuss quantitative methods to search for such effects in time-tagged photon data. Utilizing observed gamma-ray burst profiles extrapolated to GeV energies, as may expected to be observed by GLAST, we also demonstrate the extent to which these methods can be used as an empirical exploration of quantum gravity formalisms

  14. FPGA Implementation of Burst-Mode Synchronization for SOQSPK-TG

    Science.gov (United States)

    2014-06-01

    is normalized to π. The proposed burst-mode architecture is written in VHDL and verified using Modelsim. The VHDL design is implemented on a Xilinx...Document Number: SET 2014-0043 412TW-PA-14298 FPGA Implementation of Burst-Mode Synchronization for SOQSPK-TG June 2014 Final Report Test...To) 9/11 -- 8/14 4. TITLE AND SUBTITLE FPGA Implementation of Burst-Mode Synchronization for SOQSPK-TG 5a. CONTRACT NUMBER: W900KK-11-C-0032 5b

  15. Tumor Blood Vessel Dynamics

    Science.gov (United States)

    Munn, Lance

    2009-11-01

    ``Normalization'' of tumor blood vessels has shown promise to improve the efficacy of chemotherapeutics. In theory, anti-angiogenic drugs targeting endothelial VEGF signaling can improve vessel network structure and function, enhancing the transport of subsequent cytotoxic drugs to cancer cells. In practice, the effects are unpredictable, with varying levels of success. The predominant effects of anti-VEGF therapies are decreased vessel leakiness (hydraulic conductivity), decreased vessel diameters and pruning of the immature vessel network. It is thought that each of these can influence perfusion of the vessel network, inducing flow in regions that were previously sluggish or stagnant. Unfortunately, when anti-VEGF therapies affect vessel structure and function, the changes are dynamic and overlapping in time, and it has been difficult to identify a consistent and predictable normalization ``window'' during which perfusion and subsequent drug delivery is optimal. This is largely due to the non-linearity in the system, and the inability to distinguish the effects of decreased vessel leakiness from those due to network structural changes in clinical trials or animal studies. We have developed a mathematical model to calculate blood flow in complex tumor networks imaged by two-photon microscopy. The model incorporates the necessary and sufficient components for addressing the problem of normalization of tumor vasculature: i) lattice-Boltzmann calculations of the full flow field within the vasculature and within the tissue, ii) diffusion and convection of soluble species such as oxygen or drugs within vessels and the tissue domain, iii) distinct and spatially-resolved vessel hydraulic conductivities and permeabilities for each species, iv) erythrocyte particles advecting in the flow and delivering oxygen with real oxygen release kinetics, v) shear stress-mediated vascular remodeling. This model, guided by multi-parameter intravital imaging of tumor vessel structure

  16. Prevention and forecasting of rock burst hazards in coal mines

    Energy Technology Data Exchange (ETDEWEB)

    Lin-ming Dou; Cai-ping Lu; Zong-long Mu; Ming-shi Gao [China University of Mining & Technology, Xuzhou (China). State Key Laboratory for Coal Resource and Mine Safety

    2009-09-15

    Rock bursts signify extreme behavior in coal mine strata and severely threaten the safety of the lives of miners, as well as the effectiveness and productivity of miners. In our study, an elastic-plastic-brittle model for the deformation and failure of coal/rock was established through theoretical analyses, laboratory experiments and field testing, simulation and other means, which perfectly predict sudden and delayed rock bursts. Based on electromagnetic emission (EME), acoustic emission (AE) and microseismic (MS) effects in the process from deformation until impact rupture of coal-rock combination samples, a multi-parameter identification of premonitory technology was formed, largely depending on these three forms of emission. Thus a system of classification for forecasting rock bursts in space and time was established. We have presented the intensity weakening theory for rock bursts and a strong-soft-strong (3S) structural model for controlling the impact on rock surrounding roadways, with the objective of laying a theoretical foundation and establishing references for parameters for the weakening control of rock bursts. For the purpose of prevention, key technical parameters of directional hydraulic fracturing are revealed. Based on these results, as well as those from deep-hole controlled blasting in coal seams and rock, integrated control techniques were established and anti-impact hydraulic props, suitable for roadways subject to hazards from rockbursts have also been developed. These technologies have been widely used in most coal mines in China, subject to these hazards and have achieved remarkable economic and social benefits. 28 refs., 9 figs., 1 tab.

  17. Maury Journals - German Vessels

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — German vessels observations, after the 1853 Brussels Conference that set International Maritime Standards, modeled after Maury Marine Standard Observations.

  18. Observing a Burst with Sunglasses

    Science.gov (United States)

    2003-11-01

    Unique Five-Week VLT Study of the Polarisation of a Gamma-Ray Burst Afterglow "Gamma-ray bursts (GRBs)" are certainly amongst the most dramatic events known in astrophysics. These short flashes of energetic gamma-rays, first detected in the late 1960's by military satellites, last from less than one second to several minutes. GRBs have been found to be situated at extremely large ("cosmological") distances. The energy released in a few seconds during such an event is larger than that of the Sun during its entire lifetime of more than 10,000 million years. The GRBs are indeed the most powerful events since the Big Bang known in the Universe, cf. ESO PR 08/99 and ESO PR 20/00. During the past years circumstantial evidence has mounted that GRBs signal the collapse of extremely massive stars, the so-called hypernovae. This was finally demonstrated some months ago when astronomers, using the FORS instrument on ESO's Very Large Telescope (VLT), documented in unprecedented detail the changes in the spectrum of the light source ("the optical afterglow") of the gamma-ray burst GRB 030329 (cf. ESO PR 16/03). A conclusive and direct link between cosmological gamma-ray bursts and explosions of very massive stars was provided on this occasion. Gamma-Ray Burst GRB 030329 was discovered on March 29, 2003 by NASA's High Energy Transient Explorer spacecraft. Follow-up observations with the UVES spectrograph at the 8.2-m VLT KUEYEN telescope at the Paranal Observatory (Chile) showed the burst to have a redshift of 0.1685 [1]. This corresponds to a distance of about 2,650 million light-years, making GRB 030329 the second-nearest long-duration GRB ever detected. The proximity of GRB 030329 resulted in very bright afterglow emission, permitting the most extensive follow-up observations of any afterglow to date. A team of astronomers [2] led by Jochen Greiner of the Max-Planck-Institut für extraterrestrische Physik (Germany) decided to make use of this unique opportunity to study the

  19. Clinical study on the value of combining neuropsychological tests with auditory event-related potential P300 for cognitive assessment in elderly patients with cerebral small vessel disease

    Directory of Open Access Journals (Sweden)

    Xiao-ling ZHAO

    2016-11-01

    Full Text Available Objective To investigate the value of combining neuropsychological tests with auditory event-related potential (ERP P300 for cognitive assessment in elderly patients with cerebral small vessel disease (cSVD.  Methods A total of 183 elderly patients with cSVD were enrolled in this study. They were divided into 3 groups according to brain MRI: lacunar infarct (LACI group (N = 62, white matter hyperintensity (WMH group (N = 60 and LACI + WMH group (N = 61. A total of 50 brain MRI normal persons were selected as control group. Montreal Cognitive Assessment (MoCA, Chinese version was used to evaluate the cognitive function, and the amplitude and latency of P300 were measured in each group.  Results Compared with control group, the MoCA total score in LACI, WMH and LACI + WMH groups were significantly lower (P = 0.042, 0.015, 0.000, and the score in LACI + WMH group was significantly lower than that in LACI and WMH groups (P = 0.001, 0.042. In the eight cognitive domains of MoCA scale, the visual space and executive function (P = 0.006, 0.041, 0.035, delayed memory (P = 0.006, 0.012, 0.048, language (P = 0.001, 0.032, 0.047 and calculation (P = 0.009, 0.001, 0.003 in LACI + WMH group were significantly lower than those in control, LACI and WMH groups. The delayed memory in LACI group was significantly lower than that in control group (P = 0.037. The delayed memory (P = 0.005 and language (P = 0.047 in WMH group were significantly lower than those in control group. Compared with control group, the amplitudes of P300 (P = 0.025, 0.033, 0.000 in LACI, WMH and LACI + WMH groups were significantly decreased, and the latencies (P = 0.018, 0.000, 0.000 were significantly prolonged. The amplitude of P300 in LACI + WMH group was significantly lower than that in LACI and WMH groups (P = 0.041, 0.018, and the latency was significantly prolonged (P = 0.000, 0.022.  Conclusions Elderly patients of cSVD all suffer from different degrees of cognitive impairment

  20. Qualification of anticorrosive coatings in nuclear vessels

    International Nuclear Information System (INIS)

    Perez Flores, C.

    1985-01-01

    Test qualifications of the behavior of anticorrosive coating systems used in nuclear vessels in service and under the accident conditions of radiation decontamination, steam chemical resistance, thermal conductivity, weathering accelerated aging are presented and discussed. (author)

  1. Fast Fourier transformation results from gamma-ray burst profiles

    Science.gov (United States)

    Kouveliotou, Chryssa; Norris, Jay P.; Fishman, Gerald J.; Meegan, Charles A.; Wilson, Robert B.; Paciesas, W. S.

    1992-01-01

    Several gamma-ray bursts in the BATSE data have sufficiently long durations and complex temporal structures with pulses that appear to be spaced quasi-periodically. In order to test and quantify these periods we have applied fast Fourier transformations (FFT) to all these events. We have also performed cross spectral analyses of the FFT of the two extreme (high-low) energy bands in each case to determine the lead/lag of the pulses in different energies.

  2. High energy particles from {gamma}-ray bursts

    Energy Technology Data Exchange (ETDEWEB)

    Waxman, E [Weizmann Institute of Science, Rehovot (Israel)

    2001-11-15

    A review is presented of the fireball model of {gamma}-ray bursts (GRBs), and of the production in GRB fireballs of high energy protons and neutrinos. Constraints imposed on the model by recent afterglow observations, which support the association of GRB and ultra-high energy cosmic-ray (UHECR) sources, are discussed. Predictions of the GRB model for UHECR production, which can be tested with planned large area UHECR detectors and with planned high energy neutrino telescopes, are reviewed. (author)

  3. A search for spectral lines in gamma-ray bursts using TGRS

    International Nuclear Information System (INIS)

    Kurczynski, P.; Palmer, D.; Seifert, H.; Teegarden, B. J.; Gehrels, N.; Cline, T. L.; Ramaty, R.; Hurley, K.; Madden, N. W.; Pehl, R. H.

    1998-01-01

    We present the results of an ongoing search for narrow spectral lines in gamma-ray burst data. TGRS, the Transient Gamma-Ray Spectrometer aboard the Wind satellite is a high energy-resolution Ge device. Thus it is uniquely situated among the array of space-based, burst sensitive instruments to look for line features in gamma-ray burst spectra. Our search strategy adopts a two tiered approach. An automated 'quick look' scan searches spectra for statistically significant deviations from the continuum. We analyzed all possible time accumulations of spectra as well as individual spectra for each burst. Follow-up analysis of potential line candidates uses model fitting with F-test and χ 2 tests for statistical significance

  4. Ligament rupture and unstable burst behaviors of axial flaws in steam generator U-bends

    Energy Technology Data Exchange (ETDEWEB)

    Bahn, Chi Bum, E-mail: bahn@pusan.ac.kr [Pusan National University, 2 Busandaehak-ro 63 beon-gil, Geumjeong-gu, Busan 609-735 (Korea, Republic of); Oh, Young-Jin [KEPCO Engineering & Construction Co. Inc., Seongnam 463-870 (Korea, Republic of); Majumdar, Saurin [Argonne National Laboratory, Lemont, IL 60439 (United States)

    2015-11-15

    Highlights: • Ligament rupture and unstable burst pressure tests were conducted with U-bends. • In general, U-bends showed higher ligament rupture and burst pressures than straight tubes. • U-bend test data was bounded by 90% lower limit of the probabilistic models for straight tubes. • Prediction models for straight tubes could be conservatively applied to U-bends. - Abstract: Incidents of U-bend cracking in steam generator (SG) tubes have been reported, some of which have led to tube rupture. Experimental and analytical modeling efforts to determine the failure criteria of flawed SG U-bends are limited. To evaluate structural integrity of flawed U-bends, ligament rupture and unstable burst pressure tests were conducted on 57 and 152 mm bend radius U-bends with axial electrical discharge machining notches. In general, the ligament rupture and burst pressures of the U-bends were higher than those of straight tubes with similar notches. To quantitatively address the test data scatter issue, probabilistic models were introduced. All ligament rupture and burst pressures of U-bends were bounded by 90% lower limits of the probabilistic models for straight tubes. It was concluded that the prediction models for straight tubes could be applied to U-bends to conservatively evaluate the ligament rupture and burst pressures of U-bends with axial flaws.

  5. Ligament rupture and unstable burst behaviors of axial flaws in steam generator U-bends

    International Nuclear Information System (INIS)

    Bahn, Chi Bum; Oh, Young-Jin; Majumdar, Saurin

    2015-01-01

    Highlights: • Ligament rupture and unstable burst pressure tests were conducted with U-bends. • In general, U-bends showed higher ligament rupture and burst pressures than straight tubes. • U-bend test data was bounded by 90% lower limit of the probabilistic models for straight tubes. • Prediction models for straight tubes could be conservatively applied to U-bends. - Abstract: Incidents of U-bend cracking in steam generator (SG) tubes have been reported, some of which have led to tube rupture. Experimental and analytical modeling efforts to determine the failure criteria of flawed SG U-bends are limited. To evaluate structural integrity of flawed U-bends, ligament rupture and unstable burst pressure tests were conducted on 57 and 152 mm bend radius U-bends with axial electrical discharge machining notches. In general, the ligament rupture and burst pressures of the U-bends were higher than those of straight tubes with similar notches. To quantitatively address the test data scatter issue, probabilistic models were introduced. All ligament rupture and burst pressures of U-bends were bounded by 90% lower limits of the probabilistic models for straight tubes. It was concluded that the prediction models for straight tubes could be applied to U-bends to conservatively evaluate the ligament rupture and burst pressures of U-bends with axial flaws.

  6. LTP in Hippocampal Area CA1 Is Induced by Burst Stimulation over a Broad Frequency Range Centered around Delta

    Science.gov (United States)

    Grover, Lawrence M.; Kim, Eunyoung; Cooke, Jennifer D.; Holmes, William R.

    2009-01-01

    Long-term potentiation (LTP) is typically studied using either continuous high-frequency stimulation or theta burst stimulation. Previous studies emphasized the physiological relevance of theta frequency; however, synchronized hippocampal activity occurs over a broader frequency range. We therefore tested burst stimulation at intervals from 100…

  7. Method of separation of celestial gamma-ray bursts from solar flares

    International Nuclear Information System (INIS)

    Chuang, K.W.; White, R.S.; Klebesadel, R.W.; Laros, J.G.

    1991-01-01

    We recently discovered 217 ''new'' celestial gamma-ray burst candidates from the ''new'' burst search of the PVO real time data base. 1 The burst search covered the time period from September 1978 to July 1988. Sixty were confirmed by at lest on other spacecraft, e.g., ISEE-3, V-11, V-12, etc. None triggered the PVO high time resolution memory. In this paper we describe a new algorithm based ont eh relationship between time width T w and hardness ratio HR, to distinguish cosmic gamma-ray bursts from solar flares without knowing the directions of the events. The criteria for identification as a gamma-ray burst candidate are: If T ww ≤a then HR≥bT w , or T w >a then HR>c. Otherwise, the event is a solar flare candidate. Here, a, b, and c are parameter which differ for different gamma-ray burst detectors. For PVO, a=18.8 s, b=(1.38/18.8) s -1 , and c=1.38. This algorithm was tested with 83 triggered and 60 nontriggered confirmed gamma-ray burst and 30 confirmed solar flares from PVO

  8. Effects of anisotropic properties on bursting behavior of rectangular cup with a V-notch

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jeong Tai [R and D Center, TERA Co. Ltd., Seoul (Korea, Republic of); Kim, Sang Mok [R and D Center, Hyosung Power and Industrial Systems PG, Changwon (Korea, Republic of); Kang, Beom Soo [Dept. of Aerospace Engineering, Pusan National University, Busan (Korea, Republic of); Ku, Tae Wan [Engineering Research Center of Innovative Technology on Advanced Forming, Pusan National University, Busan (Korea, Republic of)

    2016-09-15

    Effects of mechanical anisotropic properties on bursting failure and its pressure of rectangular deep-drawn cup fabricated by using AA3005-H14 thin sheet are investigated to utilize for electrolyte container of lithium-ion secondary batteries. The V-notch shape with a depth of 0.1 mm and an angle of 20.0 degrees is defined on the rectangular cup, which has a thickness of 0.20 mm on the major surface and that of 0.30 mm on the minor surface. With the measured mechanical properties by uni-axial tensile tests and the defined V-notch geometry, a series of numerical prediction models considering isotropic, planar and normal anisotropic characteristics, are built-up and the bursting simulations are performed. Thereafter, the bursting fracture behavior is investigated by adopting ductile fracture criterion proposed by Cockcroft and Latham. The results predicted for the planar and the normal anisotropic models show that the bursting fracture pressure is well matched to 0.400 MPa, and the isotropic and the planar anisotropic models present a bursting fracture height of about 4.95 mm and 4.92 mm, respectively. A series of experimental investigations are undertaken to verify the bursting deformation that had been predicted. The bursting pressure and its height during experimental verifications are shown to be in good agreement with each variation of about 5.88% and roughly 0.20% with respect to the numerical results obtained using the planar anisotropic model.

  9. Swift: A gamma ray burst MIDEX

    International Nuclear Information System (INIS)

    Barthelmy, Scott

    2001-01-01

    Swift is a first of its kind multiwavelength transient observatory for gamma-ray burst astronomy. It has the optimum capabilities for the next breakthroughs in determining the origin of gamma-ray bursts and their afterglows as well as using bursts to probe the early Universe. Swift will also perform the first sensitive hard X-ray survey of the sky. The mission is being developed by an international collaboration and consists of three instruments, the Burst Alert Telescope (BAT), the X-ray Telescope (XRT), and the Ultraviolet and Optical Telescope (UVOT). The BAT, a wide-field gamma-ray detector, will detect ∼1 gamma-ray burst per day with a sensitivity 5 times that of BATSE. The sensitive narrow-field XRT and UVOT will be autonomously slewed to the burst location in 20 to 70 seconds to determine 0.3-5.0 arcsec positions and perform optical, UV, and X-ray spectrophotometry. On-board measurements of redshift will also be done for hundreds of bursts. Swift will incorporate superb, low-cost instruments using existing flight-spare hardware and designs. Strong education/public outreach and follow-up programs will help to engage the public and astronomical community. Swift has been selected by NASA for development and launch in late 2003

  10. Bursting neurons and ultrasound avoidance in crickets

    Directory of Open Access Journals (Sweden)

    Gary eMarsat

    2012-07-01

    Full Text Available Decision making in invertebrates often relies on simple neural circuits composed of only a few identified neurons. The relative simplicity of these circuits makes it possible to identify the key computation and neural properties underlying decisions. In this review, we summarize recent research on the neural basis of ultrasound avoidance in crickets, a response that allows escape from echolocating bats. The key neural property shaping behavioral output is high-frequency bursting of an identified interneuron, AN2, which carries information about ultrasound stimuli from receptor neurons to the brain. AN2's spike train consists of clusters of spikes –bursts– that may be interspersed with isolated, non-burst spikes. AN2 firing is necessary and sufficient to trigger avoidance steering but only high-rate firing, such as occurs in bursts, evokes this response. AN2 bursts are therefore at the core of the computation involved in deciding whether or not to steer away from ultrasound. Bursts in AN2 are triggered by synaptic input from nearly synchronous bursts in ultrasound receptors. Thus the population response at the very first stage of sensory processing –the auditory receptor- already differentiates the features of the stimulus that will trigger a behavioral response from those that will not. Adaptation, both intrinsic to AN2 and within ultrasound receptors, scales the burst-generating features according to the stimulus statistics, thus filtering out background noise and ensuring that bursts occur selectively in response to salient peaks in ultrasound intensity. Furthermore AN2’s sensitivity to ultrasound varies adaptively with predation pressure, through both developmental and evolutionary mechanisms. We discuss how this key relationship between bursting and the triggering of avoidance behavior is also observed in other invertebrate systems such as the avoidance of looming visual stimuli in locusts or heat avoidance in beetles.

  11. Metallurgy of steels for PWR pressure vessels

    International Nuclear Information System (INIS)

    Kepka, M.; Mocek, J.; Barackova, L.

    1980-01-01

    A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure. (B.S.)

  12. Metallurgy of steels for PWR pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Kepka, M; Mocek, J; Barackova, L [Skoda, Plzen (Czechoslovakia)

    1980-09-01

    A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure.

  13. Identification of AE Bursts by Classification of Physical and Statistical Parameters

    International Nuclear Information System (INIS)

    Mieza, J.I.; Oliveto, M.E.; Lopez Pumarega, M.I.; Armeite, M.; Ruzzante, J.E.; Piotrkowski, R.

    2005-01-01

    Physical and statistical parameters obtained with the Principal Components method, extracted from Acoustic Emission bursts coming from triaxial deformation tests were analyzed. The samples came from seamless steel tubes used in the petroleum industry and some of them were provided with a protective coating. The purpose of our work was to identify bursts originated in the breakage of the coating, from those originated in damage mechanisms in the bulk steel matrix. Analysis was performed by statistical distributions, fractal analysis and clustering methods

  14. Relativistic motion in gamma-ray bursts

    International Nuclear Information System (INIS)

    Krolik, J.H.; Pier, E.A.

    1991-01-01

    Three fundamental problems affect models of gamma-ray bursts, i.e., the energy source, the ability of high-energy photons to escape the radiation region, and the comparative weakness of X-ray emission. It is indicated that relativistic bulk motion of the gamma-ray-emitting plasma generically provides a solution to all three of these problems. Results show that, if the plasma that produces gamma-ray bursts has a bulk relativistic velocity with Lorentz factor gamma of about 10, several of the most troubling problems having to do with gamma-ray bursts are solved. 42 refs

  15. Frequency chirping during a fishbone burst

    International Nuclear Information System (INIS)

    Marchenko, V.S.; Reznik, S.N.

    2011-01-01

    It is shown that frequency chirping during fishbone activity can be attributed to the reactive torque exerted on the plasma during the instability burst, which slows down plasma rotation inside the q = 1 surface and reduces the mode frequency in the lab frame. Estimates show that the peak value of this torque can exceed the neutral beam torque in modern tokamaks. The simple line-broadened quasilinear burst model (Berk et al 1995 Nucl. Fusion 35 1661), properly adapted for the fishbone case, is capable of reproducing the key features of the bursting mode. (letter)

  16. Ballerina - pirouettes in search of gamma bursts

    DEFF Research Database (Denmark)

    Brandt, Søren Kristian; Lund, Niels; Pedersen, Henrik

    1999-01-01

    The cosmological origin of gamma ray bursts has now been established with reasonable certainty, Many more bursts will need to be studied to establish the typical distance scale, and to map out the large diversity in properties which have been indicated by the first handful of events. We are propo...... are proposing Ballerina, a small satellite to provide accurate positions and new data on the gamma-ray bursts. We anticipate a detection rate an order of magnitude larger than obtained from Beppo-SAX....

  17. Assessment of reactor vessel integrity (ARVI)

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R. E-mail: sehgal@ne.kth.se; Theerthan, A.; Giri, A.; Karbojian, A.; Willschuetz, H.G.; Kymaelaeinen, O.; Vandroux, S.; Bonnet, J.M.; Seiler, J.M.; Ikkonen, K.; Sairanen, R.; Bhandari, S.; Buerger, M.; Buck, M.; Widmann, W.; Dienstbier, J.; Techy, Z.; Kostka, P.; Taubner, R.; Theofanous, T.; Dinh, T.N

    2003-04-01

    The cost-shared project ARVI (assessment of reactor vessel integrity) involves a total of nine organisations from Europe and USA. The objective of the ARVI Project is to resolve the safety issues that remain unresolved for the melt vessel interaction phase of the in-vessel progression of a severe accident. The work consists of experiments and analysis development. Four tests were performed in the EC-FOREVER Programme, in which failure was achieved in-vessels employing the French pressure vessel steel. The tests were analysed with the commercial code ANSYS-Multiphysics, and the codes SYSTUS+ and PASULA, and quite good agreement was achieved for the failure location. Natural convection experiments in stratified pools have been performed in the SIMECO and the COPO facilities, which showed that much greater heat is transferred downwards for immiscible layers or before layers mix. A model for gap cooling and a set of simplified models for the system codes have been developed. MVITA code calculations have been performed for the Czech and Hungarian VVERs, towards evaluation of the in-vessel melt retention accident management scheme. Tests have been performed at the ULPU facility with organised flow for vessel external cooling. Considerable enhancement of the critical heat flux (CHF) was obtained. The ARVI Project has reached the halfway stage. This paper presents the results obtained thus far from the project.

  18. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    Dooley, R.A.

    1986-01-01

    This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

  19. Containment vessel drain system

    Science.gov (United States)

    Harris, Scott G.

    2018-01-30

    A system for draining a containment vessel may include a drain inlet located in a lower portion of the containment vessel. The containment vessel may be at least partially filled with a liquid, and the drain inlet may be located below a surface of the liquid. The system may further comprise an inlet located in an upper portion of the containment vessel. The inlet may be configured to insert pressurized gas into the containment vessel to form a pressurized region above the surface of the liquid, and the pressurized region may operate to apply a surface pressure that forces the liquid into the drain inlet. Additionally, a fluid separation device may be operatively connected to the drain inlet. The fluid separation device may be configured to separate the liquid from the pressurized gas that enters the drain inlet after the surface of the liquid falls below the drain inlet.

  20. Thermal Behavior of the Reactor Vessel Penetration Under External Vessel Cooling During a Severe Accident

    International Nuclear Information System (INIS)

    Kang, Kyoung-Ho; Park, Rae-Joon; Kim, Jong-Tae; Min, Byung-Tae; Lee, Ki-Young; Kim, Sang-Baik

    2004-01-01

    Experimental and analytical studies on the thermal behavior of reactor vessel penetration have been performed under external vessel cooling during a severe accident in the Korean next-generation reactor APR1400. Two types of tests, SUS-EXT and SUS-DRY with and without external vessel cooling, respectively, have been performed using sustained heating by an induction heater. Three tests have been carried out varying the cooling conditions at the vessel outer surface in the SUS-EXT tests. The experimental results have been thermally estimated using the LILAC computer code. The experimental results indicate that the inner surface of the vessel was ablated by the 45-mm thickness in the SUS-DRY test. Despite the total ablation of the welding material, the penetration was not ejected outside the vessel, which could be attributed to the thermal expansion of the penetration. Unlike the SUS-DRY test, the thickness of the ablation was ∼15 to 20 mm at most, so the welding was preserved in the SUS-EXT tests. It is concluded from the experimental results that the external vessel cooling highly affected the ablation configuration and the thermal behaviors of the vessel and the penetration. An increase in coolant mass flow rate from 0.047 to 0.152 kg/s had effects on the thermal behavior of the lower head vessel and penetration in the SUS-EXT tests. The LILAC analytical results on temperature distribution and ablation depth in the lower head vessel and penetration were very similar to the experimental results

  1. Modeling of in-vessel fission product release including fuel morphology effects for severe accident analyses

    International Nuclear Information System (INIS)

    Suh, K.Y.

    1989-10-01

    A new in-vessel fission product release model has been developed and implemented to perform best-estimate calculations of realistic source terms including fuel morphology effects. The proposed bulk mass transfer correlation determines the product of fission product release and equiaxed grain size as a function of the inverse fuel temperature. The model accounts for the fuel-cladding interaction over the temperature range between 770 K and 3000 K in the steam environment. A separate driver has been developed for the in-vessel thermal hydraulic and fission product behavior models that were developed by the Department of Energy for the Modular Accident Analysis Package (MAAP). Calculational results of these models have been compared to the results of the Power Burst Facility Severe Fuel Damage tests. The code predictions utilizing the mass transfer correlation agreed with the experimentally determined fractional release rates during the course of the heatup, power hold, and cooldown phases of the high temperature transients. Compared to such conventional literature correlations as the steam oxidation model and the NUREG-0956 correlation, the mass transfer correlation resulted in lower and less rapid releases in closer agreement with the on-line and grab sample data from the Severe Fuel Damage tests. The proposed mass transfer correlation can be applied for best-estimate calculations of fission products release from the UO 2 fuel in both nominal and severe accident conditions. 15 refs., 10 figs., 2 tabs

  2. The modulatory effect of adaptive deep brain stimulation on beta bursts in Parkinson's disease.

    Science.gov (United States)

    Tinkhauser, Gerd; Pogosyan, Alek; Little, Simon; Beudel, Martijn; Herz, Damian M; Tan, Huiling; Brown, Peter

    2017-04-01

    Adaptive deep brain stimulation uses feedback about the state of neural circuits to control stimulation rather than delivering fixed stimulation all the time, as currently performed. In patients with Parkinson's disease, elevations in beta activity (13-35 Hz) in the subthalamic nucleus have been demonstrated to correlate with clinical impairment and have provided the basis for feedback control in trials of adaptive deep brain stimulation. These pilot studies have suggested that adaptive deep brain stimulation may potentially be more effective, efficient and selective than conventional deep brain stimulation, implying mechanistic differences between the two approaches. Here we test the hypothesis that such differences arise through differential effects on the temporal dynamics of beta activity. The latter is not constantly increased in Parkinson's disease, but comes in bursts of different durations and amplitudes. We demonstrate that the amplitude of beta activity in the subthalamic nucleus increases in proportion to burst duration, consistent with progressively increasing synchronization. Effective adaptive deep brain stimulation truncated long beta bursts shifting the distribution of burst duration away from long duration with large amplitude towards short duration, lower amplitude bursts. Critically, bursts with shorter duration are negatively and bursts with longer duration positively correlated with the motor impairment off stimulation. Conventional deep brain stimulation did not change the distribution of burst durations. Although both adaptive and conventional deep brain stimulation suppressed mean beta activity amplitude compared to the unstimulated state, this was achieved by a selective effect on burst duration during adaptive deep brain stimulation, whereas conventional deep brain stimulation globally suppressed beta activity. We posit that the relatively selective effect of adaptive deep brain stimulation provides a rationale for why this approach could

  3. Treatment of measurement uncertainties at the power burst facility

    International Nuclear Information System (INIS)

    Meyer, L.C.

    1980-01-01

    The treatment of measurement uncertainty at the Power Burst Facility provides a means of improving data integrity as well as meeting standard practice reporting requirements. This is accomplished by performing the uncertainty analysis in two parts, test independent uncertainty analysis and test dependent uncertainty analysis. The test independent uncertainty analysis is performed on instrumentation used repeatedly from one test to the next, and does not have to be repeated for each test except for improved or new types of instruments. A test dependent uncertainty analysis is performed on each test based on the test independent uncertainties modified as required by test specifications, experiment fixture design, and historical performance of instruments on similar tests. The methodology for performing uncertainty analysis based on the National Bureau of Standards method is reviewed with examples applied to nuclear instrumentation

  4. The genesis of period-adding bursting without bursting-chaos in the Chay model

    International Nuclear Information System (INIS)

    Yang Zhuoqin; Lu Qishao; Li Li

    2006-01-01

    According to the period-adding firing patterns without chaos observed in neuronal experiments, the genesis of the period-adding 'fold/homoclinic' bursting sequence without bursting-chaos is explored by numerical simulation, fast/slow dynamics and bifurcation analysis of limit cycle in the neuronal Chay model. It is found that each periodic bursting, from period-1 to period-7, is separately generated by the corresponding periodic spiking pattern through two period-doubling bifurcations, except for the period-1 bursting occurring via a Hopf bifurcation. Consequently, it can be revealed that this period-adding bursting bifurcation without chaos has a compound bifurcation structure with transitions from spiking to bursting, which is closely related to period-doubling bifurcations of periodic spiking in essence

  5. The genesis of period-adding bursting without bursting-chaos in the Chay model

    International Nuclear Information System (INIS)

    Yang Zhuoqin; Lu Qishao; Li Li

    2006-01-01

    According to the period-adding firing patterns without chaos observed in neuronal experiments, the genesis of the period-adding 'fold/homoclinic' bursting sequence without bursting-chaos is explored by numerical simulation, fast/slow dynamics and bifurcation analysis of limit cycle in the neuronal Chay model. It is found that each periodic bursting, from period-1 to 7, is separately generated by the corresponding periodic spiking pattern through two period-doubling bifurcations, except for the period-1 bursting occurring via a Hopf bifurcation. Consequently, it can be revealed that this period-adding bursting bifurcation without chaos has a compound bifurcation structure with transitions from spiking to bursting, which is closely related to period-doubling bifurcations of periodic spiking in essence

  6. On Gamma-Ray Bursts

    CERN Document Server

    Ruffini, Remo; Bianco, Carlo Luciano; Caito, Letizia; Chardonnet, Pascal; Cherubini, Christian; Dainotti, Maria Giovanna; Fraschetti, Federico; Geralico, Andrea; Guida, Roberto; Patricelli, Barbara; Rotondo, Michael; Hernandez, Jorge Armando Rueda; Vereshchagin, Gregory; Xue, She-Sheng

    2008-01-01

    (Shortened) We show by example how the uncoding of Gamma-Ray Bursts (GRBs) offers unprecedented possibilities to foster new knowledge in fundamental physics and in astrophysics. After recalling some of the classic work on vacuum polarization in uniform electric fields by Klein, Sauter, Heisenberg, Euler and Schwinger, we summarize some of the efforts to observe these effects in heavy ions and high energy ion collisions. We then turn to the theory of vacuum polarization around a Kerr-Newman black hole, leading to the extraction of the blackholic energy, to the concept of dyadosphere and dyadotorus, and to the creation of an electron-positron-photon plasma. We then present a new theoretical approach encompassing the physics of neutron stars and heavy nuclei. It is shown that configurations of nuclear matter in bulk with global charge neutrality can exist on macroscopic scales and with electric fields close to the critical value near their surfaces. These configurations may represent an initial condition for the...

  7. Burst pressure and leak rate from fretted SG tubes

    International Nuclear Information System (INIS)

    Hwang, Seong Sik; Jung, Man Kyo; Kim, Hong Pyo; Kim, Joung Soo

    2005-01-01

    Steam generator(SG) tubes of a pressurized water reactor(PWR) have suffered from various types of corrosion, such as pitting, wastage and stress corrosion cracking (SCC) on both the primary and secondary side. Recently, fretting/wear degradation at the tube support region has been reported in some Korean nuclear power plants. In order to prevent the primary coolant from leaking to the secondary side, the tubes are repaired by a sleeving or plugging. It is important to establish the repair criteria to assure a reactor integrity and yet maintain the plugging ratio within the limits needed for an efficient operation. The objective of the burst test is to obtain a relationship between the burst/leak rate and the shape of the fretted flaws machined with an electro discharge machining (EDM)

  8. Complex transitions between spike, burst or chaos synchronization states in coupled neurons with coexisting bursting patterns

    International Nuclear Information System (INIS)

    Gu Hua-Guang; Chen Sheng-Gen; Li Yu-Ye

    2015-01-01

    We investigated the synchronization dynamics of a coupled neuronal system composed of two identical Chay model neurons. The Chay model showed coexisting period-1 and period-2 bursting patterns as a parameter and initial values are varied. We simulated multiple periodic and chaotic bursting patterns with non-(NS), burst phase (BS), spike phase (SS), complete (CS), and lag synchronization states. When the coexisting behavior is near period-2 bursting, the transitions of synchronization states of the coupled system follows very complex transitions that begins with transitions between BS and SS, moves to transitions between CS and SS, and to CS. Most initial values lead to the CS state of period-2 bursting while only a few lead to the CS state of period-1 bursting. When the coexisting behavior is near period-1 bursting, the transitions begin with NS, move to transitions between SS and BS, to transitions between SS and CS, and then to CS. Most initial values lead to the CS state of period-1 bursting but a few lead to the CS state of period-2 bursting. The BS was identified as chaos synchronization. The patterns for NS and transitions between BS and SS are insensitive to initial values. The patterns for transitions between CS and SS and the CS state are sensitive to them. The number of spikes per burst of non-CS bursting increases with increasing coupling strength. These results not only reveal the initial value- and parameter-dependent synchronization transitions of coupled systems with coexisting behaviors, but also facilitate interpretation of various bursting patterns and synchronization transitions generated in the nervous system with weak coupling strength. (paper)

  9. Pressurization rate effect on ligament rupture and burst pressures of cracked steam generator tubes

    International Nuclear Information System (INIS)

    Majumdar, S.; Kasza, K.

    2009-01-01

    The question of whether ligament rupture pressure or unstable burst pressure may vary significantly with pressurization rate at room temperature arose from the results of pressure tests by industry on tubes with machined part-throughwall notches. Slow (quasi-static) and fast 14 MPa/s (2000 psi/s) pressurization rate tests on specimens with nominally the same notch geometry appeared to show a significant effect of the rate of pressurization on the unstable burst pressure. Unfortunately, the slow and fast loading rate tests were conducted following two different test procedures, which could confound the results. The current series of tests were conducted on a variety of specimen geometries using a consistent test procedure to better establish the effect of pressurization rate on ligament rupture and burst pressures. (author)

  10. Pressurization rate effect on ligament rupture and burst pressures of cracked steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S.; Kasza, K. [Argonne National Laboratory, Nuclear Energy Division, Lemont, Illinois (United States)

    2009-07-01

    The question of whether ligament rupture pressure or unstable burst pressure may vary significantly with pressurization rate at room temperature arose from the results of pressure tests by industry on tubes with machined part-throughwall notches. Slow (quasi-static) and fast 14 MPa/s (2000 psi/s) pressurization rate tests on specimens with nominally the same notch geometry appeared to show a significant effect of the rate of pressurization on the unstable burst pressure. Unfortunately, the slow and fast loading rate tests were conducted following two different test procedures, which could confound the results. The current series of tests were conducted on a variety of specimen geometries using a consistent test procedure to better establish the effect of pressurization rate on ligament rupture and burst pressures. (author)

  11. Minocycline affects human neutrophil respiratory burst and transendothelial migration.

    Science.gov (United States)

    Parenti, Astrid; Indorato, Boris; Paccosi, Sara

    2017-02-01

    This study aimed at investigating the in vitro activity of minocycline and doxycycline on human polymorphonuclear (h-PMN) cell function. h-PMNs were isolated from whole venous blood of healthy subjects; PMN oxidative burst was measured by monitoring ROS-induced oxidation of luminol and transendothelial migration was studied by measuring PMN migration through a monolayer of human umbilical vein endothelial cells. Differences between multiple groups were determined by ANOVA followed by Tukey's multiple comparison test; Student's t test for unpaired data for two groups. Minocycline (1-300 µM) concentration dependently and significantly inhibited oxidative burst of h-PMNs stimulated with 100 nM fMLP. Ten micromolar concentrations, which are superimposable to C max following a standard oral dose of minocycline, promoted a 29.8 ± 4 % inhibition of respiratory burst (P minocycline impaired PMN transendothelial migration, with maximal effect at 100 µM (42.5 ± 7 %, inhibition, n = 5, P minocycline exerted on innate immune h-PMN cell function.

  12. Relativistic effects in gamma-ray bursts

    International Nuclear Information System (INIS)

    Eriksen, Erik; Groen, Oeyvind

    1999-01-01

    According to recent models of the sources of gamma-ray bursts the extremely energetic emission is caused by shells expanding with ultrarelativistic velocity. With the recent identification of optical sources at the positions of some gamma-ray bursts these ''fireball'' models have acquired an actuality that invites to use them as a motivating application when teaching special relativity. We demonstrate several relativistic effects associated with these models which are very pronounced due to the great velocity of the shell. For example a burst lasting for a month in the rest frame of an element of the shell lasts for a few seconds only, in the rest frame of our detector. It is shown how the observed properties of a burst are modified by aberration and the Doppler effect. The apparent luminosity as a function of time is calculated. Modifications due to the motion of the star away from the observer are calculated. (Author)

  13. Solar energetic particles and radio burst emission

    Directory of Open Access Journals (Sweden)

    Miteva Rositsa

    2017-01-01

    Full Text Available We present a statistical study on the observed solar radio burst emission associated with the origin of in situ detected solar energetic particles. Several proton event catalogs in the period 1996–2016 are used. At the time of appearance of the particle origin (flare and coronal mass ejection we identified radio burst signatures of types II, III and IV by inspecting dynamic radio spectral plots. The information from observatory reports is also accounted for during the analysis. The occurrence of solar radio burst signatures is evaluated within selected wavelength ranges during the solar cycle 23 and the ongoing 24. Finally, we present the burst occurrence trends with respect to the intensity of the proton events and the location of their solar origin.

  14. Bursts from the very early universe

    International Nuclear Information System (INIS)

    Silk, J.; Stodolsky, L.

    2006-01-01

    Bursts of weakly interacting particles such as neutrinos or even more weakly interacting particles such as wimps and gravitons from the very early universe would offer a much deeper 'look back time' to early epochs than is possible with photons. We consider some of the issues related to the existence of such bursts and their detectability. Characterizing the burst rate by a probability P per Hubble four-volume we find, for events in the radiation-dominated era, that the natural unit of description is the present intensity of the CMB times P. The existence of such bursts would make the observation of phenomena associated with very early times in cosmology at least conceptually possible. One might even hope to probe the transplanckian epoch if complexes more weakly interacting than the graviton can exist. Other conceivable applications include the potential detectability of the formation of 'pocket universes' in a multiverse

  15. Bursts from the very early universe

    Energy Technology Data Exchange (ETDEWEB)

    Silk, J. [Department of Physics, University of Oxford, Oxford OX1 3RH (United Kingdom); Stodolsky, L. [Max-Planck-Institut fuer Physik, Foehringer Ring 6, 80805 Munich (Germany)]. E-mail: les@mppmu.mpg.de

    2006-07-27

    Bursts of weakly interacting particles such as neutrinos or even more weakly interacting particles such as wimps and gravitons from the very early universe would offer a much deeper 'look back time' to early epochs than is possible with photons. We consider some of the issues related to the existence of such bursts and their detectability. Characterizing the burst rate by a probability P per Hubble four-volume we find, for events in the radiation-dominated era, that the natural unit of description is the present intensity of the CMB times P. The existence of such bursts would make the observation of pheno associated with very early times in cosmology at least conceptually possible. One might even hope to probe the transplanckian epoch if complexes more weakly interacting than the graviton can exist. Other conceivable applications include the potential detectability of the formation of 'pocket universes' in a multiverse.

  16. Observations of gamma-ray bursts

    International Nuclear Information System (INIS)

    Strong, I.B.; Klebesadel, R.W.; Evans, W.D.

    1975-01-01

    Observational data on gamma-ray bursts are reviewed. Information is grouped into temporal properties, energy fluxes and spectral properties, and directions and distributions of the sources in space. (BJG)

  17. Damping of type III solar radio bursts

    International Nuclear Information System (INIS)

    Levin, B.N.

    1982-01-01

    The meter- and decameter-wavelength damping of type III bursts may be attributable to stabilization of the Langmuir-wave instability of the fast-electron streams through excitation of cyclotron-branch plasma waves

  18. Optimal Codes for the Burst Erasure Channel

    Science.gov (United States)

    Hamkins, Jon

    2010-01-01

    Deep space communications over noisy channels lead to certain packets that are not decodable. These packets leave gaps, or bursts of erasures, in the data stream. Burst erasure correcting codes overcome this problem. These are forward erasure correcting codes that allow one to recover the missing gaps of data. Much of the recent work on this topic concentrated on Low-Density Parity-Check (LDPC) codes. These are more complicated to encode and decode than Single Parity Check (SPC) codes or Reed-Solomon (RS) codes, and so far have not been able to achieve the theoretical limit for burst erasure protection. A block interleaved maximum distance separable (MDS) code (e.g., an SPC or RS code) offers near-optimal burst erasure protection, in the sense that no other scheme of equal total transmission length and code rate could improve the guaranteed correctible burst erasure length by more than one symbol. The optimality does not depend on the length of the code, i.e., a short MDS code block interleaved to a given length would perform as well as a longer MDS code interleaved to the same overall length. As a result, this approach offers lower decoding complexity with better burst erasure protection compared to other recent designs for the burst erasure channel (e.g., LDPC codes). A limitation of the design is its lack of robustness to channels that have impairments other than burst erasures (e.g., additive white Gaussian noise), making its application best suited for correcting data erasures in layers above the physical layer. The efficiency of a burst erasure code is the length of its burst erasure correction capability divided by the theoretical upper limit on this length. The inefficiency is one minus the efficiency. The illustration compares the inefficiency of interleaved RS codes to Quasi-Cyclic (QC) LDPC codes, Euclidean Geometry (EG) LDPC codes, extended Irregular Repeat Accumulate (eIRA) codes, array codes, and random LDPC codes previously proposed for burst erasure

  19. Stress corrosion cracking studies of reactor pressure vessel steels. Final report

    International Nuclear Information System (INIS)

    Van Der Sluys, W.A.

    1996-10-01

    The objective of this project was to perform a critical review of the information available in open literature on stress corrosion cracking of reactor pressure vessel materials in simulated light-water-reactor (LWR) conditions, develop a test procedure for conducting stress corrosion crack growth experiments in simulated LWR environments, and conduct a test program in an effort to duplicate some of the data available from the literature. The authors concluded that stress corrosion crack growth has been observed in pressure vessel steels under laboratory test conditions. The composition of the water in most cases where growth was observed is outside of the composition specified for operating conditions. Crack growth was observed in the experiments performed in this program, and it was intermittent. The cracking would start and stop for no apparent reason. In most instances, it would not restart without the change of some external variable. In a few instances, it restarted on its own. Crack growth rates as high as 3.6 x 10 -9 m/sec were observed in pressure vessel steels in high-purity water with 8 ppm oxygen. These high crack growth rates were observed for extremely short bursts in crack extension. They could not be sustained for crack growth extensions greater than a few tenths of a millimeter. From the results of this project it appears highly unlikely that stress corrosion cracking will be observed in operating nuclear plants where the coolant composition is maintained within water chemistry guidelines. However, more work is needed to better define the contaminations that cause crack growth. The crack growth rates are so high and the threshold values for crack nucleation are so low that the conditions causing them need to be well defined and avoided

  20. Nature of gamma-ray burst sources

    International Nuclear Information System (INIS)

    Ventura, J.

    1983-01-01

    Observational evidence suggests that gamma ray bursts have a local galactic origin involving neutron stars. In this light we make a critical review of physics of the thermonuclear runaway model placing emphasis on self-consistency. We further show that some of the proposed models can be observationally excluded in the light of existing data from the Einstein Observatory. The possibility of gamma bursts arising in low mass binaries is finally discussed in the light of evolutionary scenarios leading to low luminosity systems