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Sample records for vessel head models

  1. 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)

  2. Experimental investigation of creep behavior of reactor vessel lower head

    International Nuclear Information System (INIS)

    Chu, T.Y.; Pilch, M.; Bentz, J.H.; Behbahani, A.

    1998-03-01

    The objective of the USNRC supported Lower Head Failure (LHF) Experiment Program at Sandia National Laboratories is to experimentally investigate and characterize the failure of the reactor pressure vessel (RPV) lower head due to the thermal and pressure loads of a severe accident. The experimental program is complemented by a modeling program focused on the development of a constitutive formulation for use in standard finite element structure mechanics codes. The problem is of importance because: lower head failure defines the initial conditions of all ex-vessel events; the inability of state-of-the-art models to simulate the result of the TMI-II accident (Stickler, et al. 1993); and TMI-II results suggest the possibility of in-vessel cooling, and creep deformation may be a precursor to water ingression leading to in-vessel cooling

  3. Reactor vessel head permanent shield

    International Nuclear Information System (INIS)

    Hankinson, M.F.; Leduc, R.J.; Richard, J.W.; Malandra, L.J.

    1989-01-01

    A nuclear reactor is described comprising: a nuclear reactor pressure vessel closure head; control rod drive mechanisms (CRDMs) disposed within the closure head so as to project vertically above the closure head; cooling air baffle means surrounding the control rod drive mechanisms for defining cooling air paths relative to the control rod drive mechanisms; means defined within the periphery of the closure head for accommodating fastening means for securing the closure head to its associated pressure vessel; lifting lugs fixedly secured to the closure head for facilitating lifting and lowering movements of the closure head relative to the pressure vessel; lift rods respectively operatively associated with the plurality of lifting lugs for transmitting load forces, developed during the lifting and lowering movements of the closure head, to the lifting lugs; upstanding radiation shield means interposed between the cooling air baffle means and the periphery of the enclosure head of shielding maintenance personnel operatively working upon the closure head fastening means from the effects of radiation which may emanate from the control rod drive mechanisms and the cooling air baffle means; and connecting systems respectively associated with each one of the lifting lugs and each one of the lifting rods for connecting each one of the lifting rods to a respective one of each one of the lifting lugs, and for simultaneously connecting a lower end portion of the upstanding radiation shield means to each one of the respective lifting lugs

  4. Transient stratification modelling of a corium pool in a LWR vessel lower head

    International Nuclear Information System (INIS)

    Le Tellier, R.; Saas, L.; Bajard, S.

    2015-01-01

    Highlights: • A kinetic stratification model is proposed for the simulation of the in-vessel corium behaviour during a LWR severe accident. • The different associated “modes” of vessel failure by thermal focusing effect are highlighted and discussed. • A sensitivity study for a 1650 MWe GenIII PWR is presented with this model in order to illustrate the associated R&D issues. - Abstract: In the context of light water reactor severe accidents analysis, this paper is focused on one key parameter of in-vessel corium phenomenology: the immiscible phases stratification and its impact on the heat flux distribution at the corium pool lateral boundary with the so-called focusing effect related to a “thin” top metal phase and the potential vessel failure at that point. More particularly, based on the limited knowledge of the stratification transient phenomenon derived from the MASCA-RCW experiment, a basic model is proposed that can be used for corium in lower head sensitivity analyses. It has been implemented in the PROCOR platform developed at CEA Cadarache. A short parametric study on a simple hypothetical transient is presented in order to highlight the different focusing effect “modes” that can be encountered based on this in-vessel corium pool model. An early mode may occur during the formation of the top metal layer while two other modes may appear later during the thinning of this top metal layer because of thermochemically induced mass transfers. Some associated relevant parameters (model or scenario-dependent) and modelling issues are mentioned and illustrated with some results of a Monte-Carlo based sensitivity calculation on the transient behaviour of the corium in the lower head of a 1650 MWe GenIII PWR. Within the limiting modelling hypotheses, the thermal modelling of the steel layer for small (centimetre) heights and the mass diffusivity (limited in this case to the uranium diffusivity in the oxidic layer) are main sensitive parameters

  5. 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

  6. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1984-12-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  7. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1985-09-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  8. Crack growth rates in vessel head penetration materials

    International Nuclear Information System (INIS)

    Gomez Briceno, D.; Lapena, J.; Blazquez, F.

    1994-01-01

    The cracks detected in reactor vessel head penetrations in certain European plants have been attributed to Primary Water Stress Corrosion Cracking (PWSCC). The penetrations in question are made from Inconel 600. The susceptibility of this alloy to PWSCC has been widely studied in relation to use of this material for steam generator tubes. When the first reactor vessel head penetration cracks were detected, most of the available data on crack propagation rates were from test specimens made from steam generator tubes and tested under conditions that questioned the validity of these data for assessment of the evolution of cracks in penetrations. For this reason, the scope of the Spanish Research Project on the Inspection and Repair of PWR reactor vessel head penetrations included the acquisition of data on crack propagation rates in Inconel 600, representative of the materials used for vessel head penetrations. (authors). 1 fig., 2 tabs., 6 refs

  9. Head spray nozzle in reactor pressure vessel

    International Nuclear Information System (INIS)

    Hatano, Shun-ichi.

    1990-01-01

    In a reactor pressure vessel of a BWR type reactor, a head spray nozzle is used for cooling the head of the pressure vessel and, in view of the thermal stresses, it is desirable that cooling is applied as uniformly as possible. A conventional head spray is constituted by combining full cone type nozzles. Since the sprayed water is flown down upon water spraying and the sprayed water in the vertical direction is overlapped, the flow rate distribution has a high sharpness to form a shape as having a maximum value near the center and it is difficult to obtain a uniform flow rate distribution in the circumferential direction. Then, in the present invention, flat nozzles each having a spray water cross section of laterally long shape, having less sharpness in the circumferential distribution upon spraying water to the inner wall of the pressure vessel and having a wide angle of water spray are combined, to make the flow rate distribution of spray water uniform in the inner wall of the pressure vessel. Accordingly, the pressure vessel can be cooled uniformly and thermal stresses upon cooling can be decreased. (N.H.)

  10. Estimation of center line and diameter of brain blood vessel using three-dimensional blood vessel matching method with head three-dimensional CTA image

    International Nuclear Information System (INIS)

    Maekawa, Masashi; Shinohara, Toshihiro; Nakayama, Masato; Nakasako, Noboru

    2010-01-01

    To support and automate the brain blood vessel disease diagnosis, a novel method to obtain the center line and the diameter of a blood vessel is proposed with a three-dimensional head computed tomographic angiography (CTA) image. Although the line thinning processing with distance transform or gray information is generally used to obtain the blood vessel center line, this method is not essentially one to obtain the center line and tends to yield extra lines depending on CTA images. In this study, the center line of the blood vessel is obtained by tracing the vessel. The blood vessel is traced by sequentially estimating the center point and direction of the blood vessel. The center point and direction of the blood vessel are estimated by taking the correlation between the blood vessel and a solid model of the blood vessel that is designed by considering noise influence. In addition, the vessel diameter is also estimated by correlating the blood vessel and the blood vessel model of which the diameter is variable. The validity of the proposed method is confirmed by experimentally applied the proposed method to an actual three-dimensional head CTA image. (author)

  11. Vessel head penetrations: French approach for maintenance in the PLIM program

    International Nuclear Information System (INIS)

    Champigny, F.

    2002-01-01

    Full text: In 1991, in the Bugey nuclear power plant, for the first time a leak occurred at the level of a vessel head penetration made with base nickel alloy (Inconel 600). This leak was caused by a primary stress corrosion cracking coming from inside the penetration tube. The crack was trough wall extent and primary fluid went out from the top of the vessel head. Immediately, Electricite de France launched important research programs and expertise in order to understand the root causes and propose solutions to this problem. The root causes confirmed PWSCC, and in the same time solutions for repair were studied and an inspection program was established to check the base metal of other vessel head penetrations. After several tests, repair solutions were abandoned because of their high costs (financial and dosimetry). EDF decided to replace all the vessel heads with Inconel 600 penetrations. Non destructive developments leaded to use eddy currents for detection and characterization but also televisual techniques to confirm. In a second step, in order to inspect without removing the inside thermal sleeve, eddy current and ultrasonic sword probes were achieved and used to inspect all vessel heads penetrations. Up to now, 75% of the vessel head have been replaced on the 900 MW and 1300 MW fleets but to replace wisely the last vessel heads EDF continues to perform NDE of the penetrations on the basis of safety criteria. This paper describes the different steps of the applied policy in France, NDE methods, criteria and the results obtained. (author)

  12. Primary circuit leak detection an application on PWR vessel head penetrations

    International Nuclear Information System (INIS)

    Loisy, F.; Germain, J.L.; Chauvel, L.

    1996-01-01

    In 1991, cracks were discovered and localized in the lower part of certain vessel head adapters in EDF PWR units. While awaiting the replacement of the vessel heads in question, EDF developed systems to enable continuous monitoring of vessel head penetration, by means of early detection of leaks. One of these systems in based on detection of water vapour in a confined space above the vessel head. The efficiency of the measurement chain is particularly dependent on dilution of the leakage in the confined space prior TO entry in the sampling circuit. The detection threshold for this method is on the order of 1.2 liters/hour for a dilution rate of 1500 rate of 1500 m 3 /h and a dew point of 22 deg C. This system has now been in operation on three 1300-MW PWR units for three years, and has proved to function satisfactorily. (authors)

  13. Evaluation of a pig femoral head osteonecrosis model

    Directory of Open Access Journals (Sweden)

    Kim Harry

    2010-03-01

    Full Text Available Abstract Background A major cause of osteonecrosis of the femoral head is interruption of a blood supply to the proximal femur. In order to evaluate blood circulation and pathogenetic alterations, a pig femoral head osteonecrosis model was examined to address whether ligature of the femoral neck (vasculature deprivation induces a reduction of blood circulation in the femoral head, and whether transphyseal vessels exist for communications between the epiphysis and the metaphysis. We also tested the hypothesis that the vessels surrounding the femoral neck and the ligamentum teres represent the primary source of blood flow to the femoral head. Methods Avascular osteonecrosis of the femoral head was induced in Yorkshire pigs by transecting the ligamentum teres and placing two ligatures around the femoral neck. After heparinized saline infusion and microfil perfusion via the abdominal aorta, blood circulation in the femoral head was evaluated by optical and CT imaging. Results An angiogram of the microfil casted sample allowed identification of the major blood vessels to the proximal femur including the iliac, common femoral, superficial femoral, deep femoral and circumflex arteries. Optical imaging in the femoral neck showed that a microfil stained vessel network was visible in control sections but less noticeable in necrotic sections. CT images showed a lack of microfil staining in the epiphysis. Furthermore, no transphyseal vessels were observed to link the epiphysis to the metaphysis. Conclusion Optical and CT imaging analyses revealed that in this present pig model the ligatures around the femoral neck were the primary cause of induction of avascular osteonecrosis. Since the vessels surrounding the femoral neck are comprised of the branches of the medial and the lateral femoral circumflex vessels, together with the extracapsular arterial ring and the lateral epiphyseal arteries, augmentation of blood circulation in those arteries will improve

  14. Reactor vessel closure head replacements in 1997

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    The Framatome-Jeumont Industrie consortium have completed in 1997 28 reactor vessel (RV) closure head replacements, including five on 1300 MWe class PWR units. Framatome manages the operations and handles removal and reinstallation of equipment (not including the control rod drive mechanisms (CRDM)) and the requalification tests, while JI, which manufactures the CRDMs, is involved in the CRDM cutting, re-machining and welding operations, using tools of original design, in order to optimize the RV closure head operation in terms of costs, schedule and dosage

  15. Residual life assessment of French PWR vessel head penetrations through metallurgical analysis

    International Nuclear Information System (INIS)

    Pichon, C.; Boudot, R.; Benhamou, C.; Gelpi, A.

    1994-01-01

    In September 1991, a vessel head penetration was found leaking at Bugey 3 plant during the hydrotest included in the framework of decennial In Service Inspections. Non destructive examinations performed afterwards on several other plants have shown some cracked penetrations. Destructive expertise confirmed quickly that again this new problem is related to stress corrosion cracking of Alloy 600 used as base material. During the last 15 years, similar cracking have been met in steam generator tubes and secondly in pressurizer instrumentation tubes. In spite of all the work performed since that time an extension appears to be necessary for explaining the features of this new event; however material sensitivity, stress and temperature still remain the key parameters governing the behavior of Alloy 600 in PWR environment. In this paper, only the material sensitivity of vessel head penetrations is examined through metallurgical analysis in relation with SCC tests. On the basis of vessel head field experience in combination with thermomechanical process used for fabrication of original bars criteria for a sensitivity ranking of penetrations are proposed. Metallurgical investigations and SCC tests were carried out to support this sensitivity ranking. The final aim is to use such information among those quoted above for assessment of vessel heads residual life. This document is an overview of the work performed in France concerning the material sensitivity of forged Alloy 600. It represents an important part of the assessments and investigations undertaken in France on the stress corrosion cracking phenomenon affecting the reactor vessel head penetrations in PWR's

  16. The coolability limits of a reactor pressure vessel lower head

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, T.G.; Syri, S. [Univ. of California, Santa Barbara, CA (United States)

    1995-09-01

    Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

  17. Investigating the cooling ability of reactor vessel head injection in the Maanshan PWR using CFD simulation

    International Nuclear Information System (INIS)

    Tseng Yungshin; Lin Chihhung; Wan Jongrong; Shih Chunkuan; Tsai, F. Peter

    2011-01-01

    In order to reduce the crack growth rate on the welding of penetration pipe, Pressurized Water Reactor (PWR) of Maanshan nuclear power plant (NPP) uses vessel head injection to cool vessel lid and control rod driving components. The injection flow from the cold leg is drained by the pressure difference between cold leg and upper internal components. In this study, 10 million meshes model with 4 sub-models have been developed to simulate the thermal-hydraulic behavior by commercial CFD program FLUENT. The results indicate that the injection nozzles can provide good cooling ability to reduce the maximum temperature for lid on the vessel head. The maximum temperature of vessel lid is about 293.81degC. Based on the simulated temperature, ASME CODE N-729-1 was further used to recount the effective degradation years (EDY) and reinspection years (RIY) factors. It demonstrates that the EDY and RIY factors are still less than 1.0. Therefore, the re-inspection period for Maanshan PWR would not be significantly affected by the miner temperature difference. (author)

  18. A Basic Study on the Failure of Lower Head of Nuclear Reactor Vessel by Molten Core in Severe Accident

    International Nuclear Information System (INIS)

    Cho, Jongrea; Bang, Kwanghyun; Bae, Jihoon; Kim, Changsung; Jeon, Jongwon

    2013-01-01

    This paper is analyzed by transient analysis for eight hours. Thermal conditions were carried out to interpret the data obtained from the existing experiment, and the pressures analyses were conducted considering pressure drop by applying the 1MPa. According to the analysis, a portion of the nozzle and the head is soluble, while nozzles and heads were not separated. This structural analysis has a comparative analysis of strain and displacement due to the existence of creep. Without the creep effect, strain shows 2.7% in 2D model and 4.6 % in 3D model. And, strain shows 2.9% in 2D model and 4.7 % in 3D model, in creep effect condition. Both case is satisfied to allowable strain. When comparing both analyses about creep effect, strain differences are 0.2% in 2D model and 0.1% in 3D model. Thus, it can be seen that in these analyses, the effect that creep has is minor. The purpose of this study is to develop the analysis techniques of the reactor vessel lower head under in-vessel pressure loads and thermal loads in severe accident. First, the temperature distribution in accordance with time using the thermal loads imposed on the lower head inner wall for simplified 2D model and 3D model respectively was analyzed. Second, the pressure applied on the lower head inner wall, was calculated by using the simplified 2D model and 3D model respectively. And The results of the analysis are indicated by equivalent von-mises stress and sum of the displacement, respectively. Third, the creep model and parameters used in the calculation were selected as well as the curve fitting of the experimental creep data. The plastic strain is the major cause of failure of the reactor pressure vessel. However, it can be calculated in this study that creep is not an important factor of failure of the reactor pressure vessel given the above mechanical and thermal loads

  19. Coupled thermo-mechanical analysis of corium-loaded lower head of pressure vessel

    International Nuclear Information System (INIS)

    Mishra, J.; Balasubramaniyan, V.

    2016-01-01

    A severe accident in the pressurised water reactor may lead to the relocation of core materials to the lower head of Reactor Pressure Vessel (RPV). The core debris at the bottom of RPV forms a melt pool of corium due to decay heat. The understanding of behaviour of pressure vessel, characterised by failure mode and time to failure, in this scenario is one of the important steps in predicting the accident progression. The most predominant failure mode is multi-axial creep deformation of the vessel with a non-uniform temperature field. Towards this, a numerical analysis methodology is developed for the prediction of pressure vessel deformation during the severe accidents. The methodology involves 2-D finite element modelling under multi-physics environment, which account the creep phenomena using Norton-Bailey creep law with a typical damage model of RPV material. The validation of the methodology is carried out using the results from OLHF experiment carried out in Sandia National Laboratory (SNL), USA, within the framework of an OECD. (author)

  20. Minimization of stress concentration factor in cylindrical pressure vessels with ellipsoidal heads

    International Nuclear Information System (INIS)

    Magnucki, K.; Szyc, W.; Lewinski, J.

    2002-01-01

    The paper presents the problem of stress concentration in a cylindrical pressure vessel with ellipsoidal heads subject to internal pressure. At the line, where the ellipsoidal head is adjacent to the circular cylindrical shell, a shear force and bending moment occur, disturbing the membrane stress state in the vessel. The degree of stress concentration depends on the ratio of thicknesses of both the adjacent parts of the shells and on the relative convexity of the ellipsoidal head, with the range for radius-to-thickness ratio between 75 and 125. The stress concentration was analytically described and, afterwards, the effect of these values on the stress concentration ratio was numerically examined. Results of the analysis are shown on charts

  1. Learning from EDF investigations on SG divider plates and vessel head nozzles. Evidence of prior deformation effect on stress corrosion cracking

    International Nuclear Information System (INIS)

    Deforge, D.; Duisabeau, L.; Miloudi, S.; Thebault, Y.; Couvant, T.; Vaillant, F.; Lemaire, E.

    2011-01-01

    Nickel Based alloys Stress Corrosion Cracking (SCC) has been a major concern for all the Nuclear Power Plants (NPP) utilities since the beginning of the seventies. At EDF, the nineties were marked by the occurrence of cracks on vessel head nozzles. These cracks were responsible for a leak at Bugey 3 vessel head, which was the precursor leading to the replacement of all vessel heads. From 2002, new cases of Stress Corrosion Cracking were reported on Steam Generator (SG) Divider Plates (SGDP) welded junctions. These cracks are periodically inspected inservice and reparations could be performed in case of a significant evolution of the phenomenon even if the safety issue is less relevant than for the vessel head nozzles. Both issues have led to an important non-destructive testing (NDT) program and to destructive investigations campaigns. NDT were performed on an exhaustive basis for all vessel head nozzles and for all the divider plates of 900 MWe plants. Destructive investigations were performed on more than 30 vessel head nozzles and on 6 divider plates. The last investigations were performed on samples from two decommissioned Steam Generators of Chinon B1 which present SCC cracks. In this paper, the main conclusions driven from the analysis of both NDT and destructive investigation results are reported and a comparison of the behaviours of divider plates and vessel head nozzles is given. Results give evidence that prior plastic deformation of the components before operation is fundamental for the further environmental behaviour of the material. Analysis of field experience based on parameters characteristics of prior deformation and parameters characteristics of material microstructure can be used to account for the components which are the most sensitive to SCC cracking. Some perspectives on SCC predictive models are also presented. (authors)

  2. Hygrometric measurement for on-line monitoring of PWR vessel head penetrations

    International Nuclear Information System (INIS)

    Germain, J.L.; Loisy, F.; Apolzan, S.

    1994-06-01

    In September 1991, a small leak was found on one of the reactor's upper vessel head penetrations. After inspection, other non-throughwall cracks were localized in the lower part of the vessel head adapter in questions. The same type of crack was later found inside some adapters on other French PWR units. After repairs, the safety authorities granted approval to continue unit operation, with the specific provision that a system for ongoing monitoring of the penetrations be set up. Two types of system were selected to detect leaks through any potential cracks: the first is based on nitrogen-13 detection and the second on steam detection. Both systems call for sampling the air in a confined space above the vessel head. The number and distribution of sampling taps in the circuit, and the balancing of their respective flow rates, are factors in proper monitoring of all vessel head penetrations. Gas-injection holes are also installed in the confined space. These holes are used during the sampling system qualification tests to simulate leaks in various positions and calculate the effective performance of the sampling system. Leaks are simulated using a helium-base gas tracer and measuring tracer concentrations in the sampling system. The system for measuring steam levels in air samples uses chilled-mirror hygrometers. A microcomputer takes regular readings, drives the various automatic functions of the measurement system and automatically analyses the readings so as to monitor operations and trigger an alarm at the first sign of a leak. This system has now been installed for a year and a half on three French PWR units and is functioning satisfactorily. (authors). 5 figs

  3. TMI-2 reactor-vessel head removal and damaged-core-removal planning

    International Nuclear Information System (INIS)

    Logan, J.A.; Hultman, C.W.; Lewis, T.J.

    1982-01-01

    A major milestone in the cleanup and recovery effort at TMI-2 will be the removal of the reactor vessel closure head, planum, and damaged core fuel material. The data collected during these operations will provide the nuclear power industry with valuable information on the effects of high-temperature-dissociated coolant on fuel cladding, fuel materials, fuel support structural materials, neutron absorber material, and other materials used in reactor structural support components and drive mechanisms. In addition, examination of these materials will also be used to determine accident time-temperature histories in various regions of the core. Procedures for removing the reactor vessel head and reactor core are presented

  4. 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

  5. Stress corrosion cracking in the vessel closure head penetrations of French PWR's

    International Nuclear Information System (INIS)

    Buisine, D.; Cattant, F.; Champredonde, J.; Pichon, C.; Benhamou, C.; Gelpi, A.; Vaindirlis, M.

    1994-01-01

    During a hydrotest in September 1991, part of the statutory decennial in-service inspection, a leak was detected on the vessel head of Bugey 3, which is one of the first 900 MW 3-loop PWR's in France. This leak was due to a cracked penetration used for a control rod drive mechanism. The investigations performed identified Primary Stress Corrosion Cracking of Alloy 600 as being the origin of this degradation. So a lot of the same design PWR's are a concern due to this generic problem. In this case, PWSCC was linked to: - hot temperature of the vessel head; - high residual stresses due to the welding process between peripherical penetrations and the vessel head; - sensitivity of forged Alloy 600 used for penetration manufacturing. This following paper will present the cracked analysis based, in particular, on the main results obtained in France on each of these items. These results come from the operating experience, the destructive examinations and the programs which are running on stress analysis and metallurgical characterizations. (authors). 9 figs., 2 tabs

  6. Real-time sail and heading optimization for a surface sailing vessel by extremum seeking control

    DEFF Research Database (Denmark)

    Treichel, Kai; Jouffroy, Jerome

    2010-01-01

    In this paper we develop a simplified mathematical model representing the main elements of the behaviour of sailing vessels as a basis for simulation and controller design. For adaptive real-time optimization of the sail and heading angle we then apply extremum seeking control (which is a gradient...

  7. Replacement of a vessel head, an operation which today gets easily into its stride

    International Nuclear Information System (INIS)

    Mardon, P.; Chaumont, J.C.; Lambiotte, P.

    1995-01-01

    In 1992, one year after the detection of a leak in a vessel head of the Electricite de France (EDF) Bugey 4 reactor, the head was replaced by the Framatome-Jeumont Industrie Group. Today, this group, which has developed new methods and new tools to optimize the cost, the time-delay and the dosimetry of this kind of intervention, has performed 11 additional replacements, two of which on 1300 MWe power units. This paper describes step by step the successive operations required for a complete vessel head replacement, including the testing of safety systems before starting up the reactor. (J.S.). 7 photos

  8. Effects of air vessel on water hammer in high-head pumping station

    International Nuclear Information System (INIS)

    Wang, L; Wang, F J; Zou, Z C; Li, X N; Zhang, J C

    2013-01-01

    Effects of air vessel on water hammer process in a pumping station with high-head were analyzed by using the characteristics method. The results show that the air vessel volume is the key parameter that determines the protective effect on water hammer pressure. The maximum pressure in the system declines with increasing air vessel volume. For a fixed volume of air vessel, the shape of air vessel and mounting style, such as horizontal or vertical mounting, have little effect on the water hammer. In order to obtain good protection effects, the position of air vessel should be close to the outlet of the pump. Generally, once the volume of air vessel is guaranteed, the water hammer of a entire pipeline is effectively controlled

  9. Effects of air vessel on water hammer in high-head pumping station

    Science.gov (United States)

    Wang, L.; Wang, F. J.; Zou, Z. C.; Li, X. N.; Zhang, J. C.

    2013-12-01

    Effects of air vessel on water hammer process in a pumping station with high-head were analyzed by using the characteristics method. The results show that the air vessel volume is the key parameter that determines the protective effect on water hammer pressure. The maximum pressure in the system declines with increasing air vessel volume. For a fixed volume of air vessel, the shape of air vessel and mounting style, such as horizontal or vertical mounting, have little effect on the water hammer. In order to obtain good protection effects, the position of air vessel should be close to the outlet of the pump. Generally, once the volume of air vessel is guaranteed, the water hammer of a entire pipeline is effectively controlled.

  10. Prediction of thermal margin for external cooling of reactor vessel lower head during a severe accident

    International Nuclear Information System (INIS)

    Yoon, Ho Jun; Suh, Kune Y.

    1998-01-01

    In the TMI-2 accident, approximately nineteen (19) tons of molten core material drained into the lower plenum. One of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 .deg. C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident management strategies. As an advanced in-vessel design concept, the COrium Attak Syndrome Immunization Structures (COASIS) are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in -vessel (COASISI) and ex-vessel (COASISO) were demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the TMI-2 and the Korean Standard Nuclear Power Plant (KSNPP) reactors. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. In studying the in-vessel severe accident phenomena, one of the main goals is to verify the cooling mechanism in the reactor vessel lower plenum and thereby to prevent the vessel failure from thermal attack by the molten debris. This paper presents the first-principle calculation results for the thermal margin for the case of external cooling of the reactor vessel lower head. Adopting the method presented by F.B. Cheung, et al., we calculated the departure from nucleate boiling ratio (DNBR) for the three cases of pool boiling, flow boiling

  11. Experimental Investigation of Creep Behavior of Reactor Vessel Lower Head

    International Nuclear Information System (INIS)

    Chu, T.Y.; Pilch, M.; Bentz, J.H.; Behbahani, A.

    1999-01-01

    The authors report a study which aimed at experimentally and numerically investigating and characterizing the failure of a reactor pressure vessel (RPV) lower head due to thermal and pressure loads generated by a severe accident. They present the experimental apparatus which is based on a scaled version of the lower part of a TMI-like reactor pressure vessel without vessel skirt. They report and comment the results obtained during the first five experiments: uniform heating and non penetrations, centre-peaked heat flux and no penetrations, edge-peaked heat flux and no penetrations, uniform heating with penetrations, edge-peaked heat flux with penetrations. They compare the third and fifth experience (those with edge-peaked heat flux)

  12. A preliminary assessment of the effects of heat flux distribution and penetration on the creep rupture of a reactor vessel lower head

    International Nuclear Information System (INIS)

    Chu, T.Y.; Bentz, J.; Simpson, R.; Witt, R.

    1997-01-01

    The objective of the Lower Head Failure (LHF) Experiment Program is to experimentally investigate and characterize the failure of the reactor vessel lower head due to thermal and pressure loads under severe accident conditions. The experiment is performed using 1/5-scale models of a typical PWR pressure vessel. Experiments are performed for various internal pressure and imposed heat flux distributions with and without instrumentation guide tube penetrations. The experimental program is complemented by a modest modeling program based on the application of vessel creep rupture codes developed in the TMI Vessel Investigation Project. The first three experiments under the LHF program investigated the creep rupture of simulated reactor pressure vessels without penetrations. The heat flux distributions for the three experiments are uniform (LHF-1), center-peaked (LHF-2), and side-peaked (LHF-3), respectively. For all the experiments, appreciable vessel deformation was observed to initiate at vessel wall temperatures above 900K and the vessel typically failed at approximately 1000K. The size of failure was always observed to be smaller than the heated region. For experiments with non-uniform heat flux distributions, failure typically occurs in the region of peak temperature. A brief discussion of the effect of penetration is also presented

  13. Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head

    International Nuclear Information System (INIS)

    Villanueva, Walter; Tran, Chi-Thanh; Kudinov, Pavel

    2012-01-01

    Highlights: ► We consider a severe accident in a BWR with melt pool formation in the lower head. ► We study the influence of pool depth on vessel failure mode with creep analysis. ► There are two modes of failure; ballooning of vessel bottom and a localized creep. ► External vessel cooling can suppress creep and subsequently prevent vessel failure. - Abstract: In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed and then melt pool formation. Nordic BWRs rely on reactor cavity flooding as a means for ex-vessel melt coolability and ultimate termination of the accident progression. However, different modes of vessel failure may result in different regimes of melt release from the vessel, which determine initial conditions for melt coolant interaction and eventually coolability of the debris bed. The goal of this study is to define if retention of decay-heated melt inside the reactor pressure vessel is possible and investigate modes of the vessel wall failure otherwise. The mode of failure is contingent upon the ultimate mechanical strength of the vessel structures under given mechanical and thermal loads and applied cooling measures. The influence of pool depth and respective transient thermal loads on the reactor vessel failure mode is studied with coupled thermo-mechanical creep analysis. Efficacy of control rod guide tube (CRGT) cooling and external vessel wall cooling as potential severe accident management measures is investigated. First, only CRGT cooling is considered in simulations revealing two different modes of vessel failure: (i) a ‘ballooning’ of the vessel bottom and (ii) a ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool. Second, possibility of in-vessel retention with CRGT and external vessel cooling is investigated. We found that the external vessel

  14. In-vessel retention modeling capabilities in MAAP5

    International Nuclear Information System (INIS)

    Paik, Chan Y.; Lee, Sung Jin; Zhou, Quan; Luangdilok, W.; Reeves, R.W.; Henry, R.E.; Plys, M.; Scobel, J.H.

    2012-01-01

    Modular Accident Analysis Program (MAAP) is an integrated severe accident analysis code for both light water and heavy water reactors. New and improved models to address the complex phenomena associated with in-vessel retention (IVR) were incorporated into MAAP5.01. They include: -a) time-dependent volatile and non-volatile decay heat, -b) material properties at high temperatures, -c) finer vessel wall nodalization, -d) new correlations for natural convection heat transfer in the oxidic pool, -e) refined metal layer heat transfer to the reactor vessel wall and surroundings, -f) formation of a heavy metal layer, and -g) insulation cooling channel model and associated ex-vessel heat transfer and critical heat flux correlations. In this paper, the new and improved models in MAAP5.01 are described and sample calculation results are presented for the AP1000 passive plant. For the IVR evaluation, a transient calculation is useful because the timing of corium relocation, decaying heat load, and formation of separate layers in the lower plenum all affect integrity of the lower head. The key parameters affecting the IVR success are the metal layer emissivity and thickness of the top metal layer, which depends on the amount of steel in the oxidic pool and in the heavy metal layer. With the best estimate inputs for the debris mixing parameters in a conservative IVR scenario, the AP1000 plant results show that the maximum ex-vessel heat flux to CHF ratio is about 0.7, which occurs before 10.000 seconds when the decay heat is high. The AP1000 plant results demonstrate how MAAP5.01 can be used to evaluate IVR and to gain insight into responses of the lower head during a severe accident

  15. Short and long term maintenance strategy for reactor vessel head penetrations

    International Nuclear Information System (INIS)

    Teissier, A.; Heuze, A.

    1995-01-01

    This paper presents elements based on : surveys, operating inspection, theoretical studies, safety analysis, laboratory results, that enabled to determine maintenance options and short and long term strategies for processing on reactor vessel head leaks. (TEC). 1 tab

  16. Evaluation of the stress distribution on the pressure vessel head with multi-openings

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K.S.; Kim, T.W.; Jeong, K.H.; Lee, G.M. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    This report discusses and analyzes the stress distribution on the pressure vessel head with multi-openings(3 PSV nozzles, 2 SDS nozzles and 1 Man Way) according to patterns of the opening distance. The pressurizer of Korea Standardized Nuclear Power Plant(Ulchin 3 and 4), which meets requirements of the cyclic operation and opening design defined by ASME code, was used as the basic model for that. Stress changes according to the distance between openings were investigated and the factors which should be considered for the opening design were analyzed. Also, the nozzle loads at Level A, B conditions and internal pressure were applied in order to evaluate changes of head stress distributions due to nozzle loads. (author). 6 refs., 29 figs., 4 tabs.

  17. Studies on core melt behaviour in a BWR pressure vessel lower head

    International Nuclear Information System (INIS)

    Lindholm, I.; Ikonen, K.; Hedberg, K.

    1999-01-01

    Core debris behaviour in the Nordic BWR lower head was investigated numerically using MELCOR and MAAP4 codes. Lower head failure due to penetration failure was studied with more detailed PASULA code taking thermal boundary conditions from MELCOR calculations. Creep rupture failure mode was examined with the two integral codes. Also, the possibility to prevent vessel failure by late reflooding was assessed in this study. (authors)

  18. Thermal and stress analyses of the reactor pressure vessel lower head of the Three Mile Island Unit 2

    International Nuclear Information System (INIS)

    Hashimoto, K.; Onizawa, K.; Kurihara, R.; Kawasaki, S.; Soda, K.

    1992-01-01

    Thermal and stress analyses were performed using the finite element analysis code ABAQUS to clarify the factors which caused tears in the stainless steel liner of the reactor pressure vessel lower head of the Three Mile Island Unit 2 (TMI-2) reactor pressure vessel during the accident on 28 March 1979. The present analyses covered the events which occurred after approximately 20 tons of molten core material were relocated to the lower head of the reactor pressure vessel. They showed that the tensile stress was highest in the case where the relocated core material consisting of homogeneous UO 2 debris was assumed to attack the lower head and the debris was then quenched. The peak tensile stress was in the vicinity of the welded zone of the penetration nozzle. This result agrees with the findings from the examination of the TMI-2 reactor pressure vessel that major tears in the stainless steel liner were observed around two penetration nozzles of the lower head. (author)

  19. Development of automated ultrasonic device for in-service inspection of ABWR pressure vessel bottom head

    International Nuclear Information System (INIS)

    Kojima, Y.; Matsuyama, A.

    1995-01-01

    An automated device and its controller have been developed for the bottom head weld examination of pressure vessel of Advanced Boiling Water Reactor (ABWR). The internal pump casings and the housings of control rod prevent a conventional ultrasonic device from scanning the required inspection zone. With this reason, it is required to develop a new device to examine the bottom head area of ABWR. The developed device is characterized by the following features. (1) Composed of a mother vehicle and a compact inspection vehicle. They are connected only by an electric wire without using the conventional arm mechanism. (2) The mother vehicle travels on a track and lift up the inspection vehicle to the vessel. (3) The mother vehicle can automatically attach the inspection vehicle to the bottom head, and detach the inspection vehicle from it. (4) Collision avoidance control function with a touch sensor is installed at the front of the inspection vehicle. The device was successfully demonstrated using a mock-up of reactor pressure vessel

  20. Residual stresses of manufacture on the PWR vessel head penetrations

    International Nuclear Information System (INIS)

    Le Hong, S.; Todeschini, P.; Ternon, F.; Cipiere, M.F.; Gimond, C.; Faure, F.

    1997-01-01

    Since the detection in September 1991 of a leakage on a vessel head adapter of Bugey 3 during the decadal hydro-test, a study has been led by Framatome and EDF on the phenomenon, which has been identified as a stress corrosion cracking. The stress parameter particularly is an important factor in regard to the behaviour of the Alloy 600 in primary water. It has been the subject of a calculation program, which is not presented here, and of an experimental program which contents: 1 - the determination of residual stresses on the inner surface of the adapter and on the weld metal by the hole method and the diffraction of X-Rays on representative mock-ups and on a vessel head during manufacturing; 2 - the visualization of the stress field at the surface by corrosion tests on representative mock-ups in sodium hydroxide at 350 o C. The results are globally consistent with each other and give an important contribution to the interpretation of the results of the controls on site. (authors)

  1. Galectin-1 Inhibitor OTX008 Induces Tumor Vessel Normalization and Tumor Growth Inhibition in Human Head and Neck Squamous Cell Carcinoma Models.

    Science.gov (United States)

    Koonce, Nathan A; Griffin, Robert J; Dings, Ruud P M

    2017-12-09

    Galectin-1 is a hypoxia-regulated protein and a prognostic marker in head and neck squamous cell carcinomas (HNSCC). Here we assessed the ability of non-peptidic galectin-1 inhibitor OTX008 to improve tumor oxygenation levels via tumor vessel normalization as well as tumor growth inhibition in two human HNSCC tumor models, the human laryngeal squamous carcinoma SQ20B and the human epithelial type 2 HEp-2. Tumor-bearing mice were treated with OTX008, Anginex, or Avastin and oxygen levels were determined by fiber-optics and molecular marker pimonidazole binding. Immuno-fluorescence was used to determine vessel normalization status. Continued OTX008 treatment caused a transient reoxygenation in SQ20B tumors peaking on day 14, while a steady increase in tumor oxygenation was observed over 21 days in the HEp-2 model. A >50% decrease in immunohistochemical staining for tumor hypoxia verified the oxygenation data measured using a partial pressure of oxygen (pO₂) probe. Additionally, OTX008 induced tumor vessel normalization as tumor pericyte coverage increased by approximately 40% without inducing any toxicity. Moreover, OTX008 inhibited tumor growth as effectively as Anginex and Avastin, except in the HEp-2 model where Avastin was found to suspend tumor growth. Galectin-1 inhibitor OTX008 transiently increased overall tumor oxygenation via vessel normalization to various degrees in both HNSCC models. These findings suggest that targeting galectin-1-e.g., by OTX008-may be an effective approach to treat cancer patients as stand-alone therapy or in combination with other standards of care.

  2. The Effective Convectivity Model for Simulation and Analysis of Melt Pool Heat Transfer in a Light Water Reactor Pressure Vessel Lower Head

    International Nuclear Information System (INIS)

    Tran, Chi Thanh

    2009-09-01

    Severe accidents in a Light Water Reactor (LWR) have been a subject of intense research for the last three decades. The research in this area aims to reach understanding of the inherent physical phenomena and reduce the uncertainties in their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors, and to evaluation of the proposed accident management schemes for mitigating the consequences of severe accidents. In a hypothetical severe accident there is likelihood that the core materials will be relocated to the lower plenum and form a decay-heated debris bed (debris cake) or a melt pool. Interactions of core debris or melt with the reactor structures depend to a large extent on the debris bed or melt pool thermal hydraulics. In case of inadequate cooling, the excessive heat would drive the structures' overheating and ablation, and hence govern the vessel failure mode and timing. In turn, threats to containment integrity associated with potential ex-vessel steam explosions and ex-vessel debris uncoolability depend on the composition, superheat, and amount of molten corium available for discharge upon the vessel failure. That is why predictions of transient melt pool heat transfer in the reactor lower head, subsequent vessel failure modes and melt characteristics upon the discharge are of paramount importance for plant safety assessment. The main purpose of the present study is to develop a method for reliable prediction of melt pool thermal hydraulics, namely to establish a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident. To achieve the goal, an approach to efficient use of Computational Fluid Dynamics (CFD) has been proposed to guide and support the development of models suitable for accident analysis. The CFD method, on the one hand, is

  3. The retrograde transverse cervical artery as a recipient vessel for free tissue transfer in complex head and neck reconstruction with a vessel-depleted neck.

    Science.gov (United States)

    Ciudad, Pedro; Agko, Mouchammed; Manrique, Oscar J; Date, Shivprasad; Kiranantawat, Kidakorn; Chang, Wei Ling; Nicoli, Fabio; Lo Torto, Federico; Maruccia, Michele; Orfaniotis, Georgios; Chen, Hung-Chi

    2017-11-01

    Reconstruction in a vessel-depleted neck is challenging. The success rates can be markedly decreased because of unavailability of suitable recipient vessels. In order to obtain a reliable flow, recipient vessels away from the zone of fibrosis, radiation, or infection need to be explored. The aim of this report is to present our experience and clinical outcomes using the retrograde flow coming from the distal transverse cervical artery (TCA) as a source for arterial inflow for complex head and neck reconstruction in patients with a vessel-depleted neck. Between July 2010 and June 2016, nine patients with a vessel-depleted neck underwent secondary head and neck reconstruction using the retrograde TCA as recipient vessel for microanastomosis. The mean age was 49.6 years (range, 36 to 68 years). All patients had previous bilateral neck dissections and all, except one, had also received radiotherapy. Indications included neck contracture release (n = 3), oral (n = 1), mandibular (n = 3) and pharyngoesophageal (n = 2) reconstruction necessitating free anterolateral thigh (n = 3) and medial sural artery (n = 1) perforator flaps, fibula (n = 3) and ileocolon (n = 2) flaps respectively. There was 100% flap survival rate with no re-exploration or any partial flap loss. One case of intra-operative arterial vasospasm at the anastomotic suture line was managed intra-operatively with vein graft interposition. There were no other complications or donor site morbidity during the follow-up period. In a vessel-depleted neck, the reverse flow of the TCA may be a reliable option for complex secondary head and neck reconstruction in selected patients. © 2017 Wiley Periodicals, Inc.

  4. INETEC new system for inspection of PWR reactor pressure vessel head

    International Nuclear Information System (INIS)

    Nadinic, B.; Postruzin, Z.

    2004-01-01

    INETEC Institute for Nuclear Technology developed new equipment for inspection of PWR and VVER reactor pressure vessel head. The new advances in inspection technology are presented in this article, as the following: New advance manipulator for inspection of RPVH with high speed of inspection possibilities and total automated work; New sophisticated software for manipulator driving which includes 3D virtual presentation of manipulator movement and collision detection possibilities; New multi axis controller MAC-8; New end effector system for inspection of penetration tube and G weld; New eddy current and ultrasonic probes for inspection of G weld and penetration tube; New Eddy One Raster scan software for analysis of eddy current data with mant advanced features which allows easy and quick data analysis. Also the results of laboratory testing and laboratory qualification are presented on reactor pressure vessel head mock, as well as obtained speed of inspection and quality of collected data.(author)

  5. Feasibility study for CPR1000 incore measurement instrumentation educed from the reactor pressure vessel upper head

    International Nuclear Information System (INIS)

    Guang Jianwei; Liu Qian; Li Wenhong; Duan Yuangang

    2010-01-01

    The article discusses about the feasibility of in-core measurement instrumentation educed from the reactor pressure vessel (RPV) upper head. Incore instrumentation educed from the reactor pressure vessel upper head is one of advanced technology in the third generation nuclear power plant. This technology can reduce the manufacture problem of RPV; decrease the manufacture time effectively. Furthermore, this technology can get rid of the trouble for loss of water caused by many penetrations in the RPV bottom head, can increase security of nuclear power plant. By the description of structure analysis, comparison, maturity for four type incore instrumentation detectors, the incore instrumentation can be educed from RPV upper head, which can increase reactor's security, reduce the manufacture time, decrease group dose in refueling period. The core design ability can be enhanced through this study. (authors)

  6. In-vessel core debris retention through external flooding of the reactor pressure vessel. SCDAP/RELAP5 assessment for the SBWR lower head

    International Nuclear Information System (INIS)

    Heel, A.M.J.M. van.

    1995-09-01

    In this report the results are discussed from various analyses on the feasibility and phenomenology of the External Flooding (EF) concept for an SBWR lower head, filled with a large heat generating corium mass. In applying External Flooding as an accident management strategy after or during core melt down, the lower drywell is filled with water up to a level where a large portion of the Reactor Pressure Vessel (RPV) is flooded. The purpose of this method is to establish cooling of the vessel wall, that is challenged by the heat load resulting from the corium, in such a way that its structural integrity if not endangered. The analysis discussed in this report focus on the thermal response of the vessel wall and the ex-vessel boiling processes under the conditions described above. For these analyses the SCDAP/REALP5 MOD 3.1 code was used. The major outcome of the calculations is, that a major part of the vessel wall remains well below themelting temperature of carbon steel, as long as flooding of the external surface of the lower head is established. The SCDAP/RELAP5 analyses indicated that low-quality Critical Heat Flux (CHF) was not exceeded, under all the conditions that had been tested. However, a comaprison of the heat fluxes, as calculated in RELAP5, with the CHF values obtained from the Zuber correlation and the Vishnev correction factor (for boiling at inclined surfaces) proved that CHF values, based on these criteria, were exceeded in several surface points of the lower head mesh. The correlations, as resident in the current version of RELAP5 MOD 3.1, might lead to over-estimation of CHF for the EF analyses discussed in this report. The use of the more conservative Zuber correlation with the Vishnev correction factor is recommended for EF analyses. (orig.)

  7. Validation of ASTEC V2 models for the behaviour of corium in the vessel lower head

    International Nuclear Information System (INIS)

    Carénini, L.; Fleurot, J.; Fichot, F.

    2014-01-01

    The paper is devoted to the presentation of validation cases carried out for the models describing the corium behaviour in the “lower plenum” of the reactor vessel implemented in the V2.0 version of the ASTEC integral code, jointly developed by IRSN (France) and GRS (Germany). In the ASTEC architecture, these models are grouped within the single ICARE module and they are all activated in typical accident scenarios. Therefore, it is important to check the validity of each individual model, as long as experiments are available for which a single physical process is involved. The results of ASTEC applications against the following experiments are presented: FARO (corium jet fragmentation), LIVE (heat transfer between a molten pool and the vessel), MASCA (separation and stratification of corium non miscible phases) and OLHF (mechanical failure of the vessel). Compared to the previous ASTEC V1.3 version, the validation matrix is extended. This work allows determining recommended values for some model parameters (e.g. debris particle size in the fragmentation model and criterion for debris bed liquefaction). Almost all the processes governing the corium behaviour, its thermal interaction with the vessel wall and the vessel failure are modelled in ASTEC and these models have been assessed individually with satisfactory results. The main uncertainties appear to be related to the calculation of transient evolutions

  8. Modeling of heat and mass transfer processes during core melt discharge from a reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Bui, V.A.; Nourgaliev, R.R. [Royal Institute of Technology, Stockholm (Sweden)] [and others

    1995-09-01

    The objective of the paper is to study heat and mass transfer processes related to core melt discharge from a reactor vessel is a severe light water reactor accident. The phenomenology of the issue includes (1) melt convection in and heat transfer from the melt pool in contact with the vessel lower head wall; (2) fluid dynamics and heat transfer of the melt flow in the growing discharge hole; and (3) multi-dimensional heat conduction in the ablating lower head wall. A program of model development, validation and application is underway (i) to analyse the dominant physical mechanisms determining characteristics of the lower head ablation process; (ii) to develop and validate efficient analytic/computational methods for estimating heat and mass transfer under phase-change conditions in irregular moving-boundary domains; and (iii) to investigate numerically the melt discharge phenomena in a reactor-scale situation, and, in particular, the sensitivity of the melt discharge transient to structural differences and various in-vessel melt progression scenarios. The paper presents recent results of the analysis and model development work supporting the simulant melt-structure interaction experiments.

  9. Manufacturing and properties of closure head forging integrated with flange for PWR reactor pressure vessel

    International Nuclear Information System (INIS)

    Tomoharu Sasaki; Iku Kurihara; Etsuo Murai; Yasuhiko Tanaka; Koumei Suzuki

    2003-01-01

    Closure head forging (SA508, Gr.3 Cl.1) integrated with flange for PWR reactor pressure vessel has been developed. This is intended to enhance structural integrity of closure head resulted in elimination of ISI, by eliminating weld joint between closure head and flange in the conventional design. Manufacturing procedures have been established so that homogeneity and isotropy of the material properties can be assured in the closure head forging integrated with flange. Acceptance tensile and impact test specimens are taken and tested regarding the closure head forging integrated with flange as very thick and complex forgings. This paper describes the manufacturing technologies and material properties of the closure head forging integrated with flange. (orig.)

  10. In-vessel inspection before head removal: TMI II: Phase I. (Conceptual development)

    International Nuclear Information System (INIS)

    Calloway, N.E.; Greenlee, D.W.; Lawrence, G.R.; Paglia, A.L.; Piatt, T.D.; Tucker, B.A.

    1981-08-01

    The objective of the task is to provide for an internal inspection of the reactor vessel and the fuel assemblies prior to reactor vessel head removal. Because the degree of damage to equipment and fuel in the TMI-II reactor is not precisely known, it is important that as much information as possible be obtained on present conditions inside the reactor. This information will serve to benchmark the various analyses already completed or underway and will also guide the development of programs to obtain more information on the TMI-II core damage. In addition, the early look will provide data for planning the reactor disassembly program

  11. 3-D segmentation of retinal blood vessels in spectral-domain OCT volumes of the optic nerve head

    Science.gov (United States)

    Lee, Kyungmoo; Abràmoff, Michael D.; Niemeijer, Meindert; Garvin, Mona K.; Sonka, Milan

    2010-03-01

    Segmentation of retinal blood vessels can provide important information for detecting and tracking retinal vascular diseases including diabetic retinopathy, arterial hypertension, arteriosclerosis and retinopathy of prematurity (ROP). Many studies on 2-D segmentation of retinal blood vessels from a variety of medical images have been performed. However, 3-D segmentation of retinal blood vessels from spectral-domain optical coherence tomography (OCT) volumes, which is capable of providing geometrically accurate vessel models, to the best of our knowledge, has not been previously studied. The purpose of this study is to develop and evaluate a method that can automatically detect 3-D retinal blood vessels from spectral-domain OCT scans centered on the optic nerve head (ONH). The proposed method utilized a fast multiscale 3-D graph search to segment retinal surfaces as well as a triangular mesh-based 3-D graph search to detect retinal blood vessels. An experiment on 30 ONH-centered OCT scans (15 right eye scans and 15 left eye scans) from 15 subjects was performed, and the mean unsigned error in 3-D of the computer segmentations compared with the independent standard obtained from a retinal specialist was 3.4 +/- 2.5 voxels (0.10 +/- 0.07 mm).

  12. 1D/2D analyses of the lower head vessel in contact with high temperature melt

    International Nuclear Information System (INIS)

    Chang, Jong Eun; Cho, Jae Seon; Suh, Kune Y.; Chung, Chang H.

    1998-01-01

    One- and two-dimensional analyses were performed for the ceramic/metal melt and the vessel to interpret the temperature history of the outer surface of the vessel wall measured from typical Al 2 O 3 /Fe thermite melt tests LAVA (Lower-plenum Arrested Vessel Attack) spanning heatup and cooldown periods. The LAVA tests were conducted at the Korea Atomic Energy Research Institute (KAERI) during the process of high temperature molten material relocation from the delivery duct down into the water in the test vessel pressurized to 2.0 MPa. Both analyses demonstrated reasonable predictions of the temperature history of the LHV (Lower Head Vessel). The comparison sheds light on the thermal hydraulic and material behavior of the high temperature melt within the hemispherical vessel

  13. Critical heat flux for APR1400 lower head vessel during a severe accident

    International Nuclear Information System (INIS)

    Noh, Sang W.; Suh, Kune Y.

    2013-01-01

    Highlights: ► Studied boiling on downward-facing hemispherical vessel with asymmetric thermal insulator. ► Scaled the APR1400 lower head linearly down by 1/10 including ICI tubes and shear keys. ► Performed thermal analysis using ANSYS V11.0 to determine the internal temperature and heat flux. ► Performed tests to obtain the CHF with saturated demineralized water at atmospheric pressure. ► Measured CHF accounting for 3D random flow effect expected in the APR1400 application. -- Abstract: Corium Ablation Stopper Apparatus (CASA) has a downward-facing hemispherical vessel and geometrically asymmetric thermal insulator of the Advanced Power Reactor 1400 MWe (APR1400) scaled linearly down by 1/10, as well as sixty-one in-core instrumentation (ICI) tubes and four shear keys. The heated vessel plays a pivotal role in CASA depending on the configuration of the oxide pool and metal layer to bring about the focusing effect expected of a molten pool in the lower head during a severe accident. The heated vessel was designed through a trial-and-error method and thermal analysis. Thermal analysis was performed using ANSYS V11.0 to investigate the effect of the internal temperature and heat flux on the integral hemispherical copper vessel. The CASA tests were carried out to obtain the critical heat flux (CHF) with saturated and demineralized water at the atmospheric pressure (0.1 MPa). The CHF in the metal layer through the hemispherical channel was found to be lower than that in the ULPU-2400 configuration V data through the streamlined thermal insulator. The experimental CHF was measured and obtained through the CASA hemispherical heated surface accounting for the three-dimensional random flow effect expected in the APR1400 application

  14. Mechanical properties and examination of cracking in TMI-2 pressure vessel lower head material

    International Nuclear Information System (INIS)

    Diercks, D.R.; Neimark, L.A.

    1993-09-01

    Mechanical tests have been conducted on material from 15 samples removed from the lower head of the Three Mile Island unit 2 nuclear reactor pressure vessel. Measured properties include tensile properties and hardness profiles at room temperature, tensile and creep properties at temperatures of 600 to 1200 degrees C, and Charpy V-notch impact properties at -20 to +300 degrees C. These data, which were used in the subsequent analyses of the margin-to-failure of the lower head during the accident, are presented here. In addition, the results of metallographic and scanning electron microscope examinations of cladding cracking in three of the lower head samples are discussed

  15. Application of directional solidification ingot (LSD) in forging of PWR reactor vessel heads

    International Nuclear Information System (INIS)

    Benhamou, C.; Poitrault, I.

    1985-09-01

    Creusot-Loire Industrie uses this type of ingot for manufacture of Framatome 1300 and 1450 MW 4-loop PWR reactor vessel heads. This type of ingot offers a number advantages: improved internal soundness; greater chemical, structural and mechanical homogeneity of the finished part; simplified forging process. After a brief description of the pouring and solidification processes, this paper presents an analysis of the results of examinations performed on the prototype forging, as well as review of results obtained during industrial fabrication of dished heads from LSD ingots. The advantages of the LSD ingot over conventional ingots are discussed in conclusion

  16. Modelling of RPV lower head under core melt severe accident condition using OpenFOAM

    International Nuclear Information System (INIS)

    Madokoro, Hiroshi; Kretzschmar, Frank; Miassoedov, Alexei

    2017-01-01

    Although six years have been passed since the tragic severe accident at Fukushima Daiichi, still large uncertainties exist in modeling of core degradation and reactor pressure vessel (RPV) failure. It is extremely important to obtain a better understanding of complex phenomena in the lower head in order to improve accident management measures. The possible failure mode of reactor pressure vessel and its failure time are especially a matter of importance. Thermal behavior of the molten pool can be simulated by the Phase-change Effective Convectivity Model (PECM), which is a distributed-parameter model developed in the Royal Institute of Technology (KTH), Sweden. The model calculates convective currents not using a pure CFD approach but based on so called “characteristic velocities” that are determined by empirical correlations depending on the geometry and physical properties of the molten pool. At the Karlsruhe Institute of Technology (KIT), the PECM has been implemented in the open-source CFD software OpenFOAM in order to receive detailed predictions of a core melt behavior in the RPV lower head under severe accident conditions. An advantage of using OpenFOAM is that it is very flexible to add and modify models and physical properties. In the current work, the solver is extended to couple PECM with a structure analysis model of the vessel wall. The model considers thermal expansion, plasticity, creep and damage. The model and physical properties are based on those implemented in ANSYS. Although the previous implementation had restriction that the amount of and geometry of the melt cannot be changed, our coupled model allows flexibility of the melt amount and geometry. The extended solver was used to simulate the LIVE-L1 and -L7V experiments and has demonstrated good prediction of the temperature distribution in the molten pool and heat flux distribution through the vessel wall. Regarding the vessel failure the model was applied to one of the FOREVER tests

  17. Blood Vessel Normalization in the Hamster Oral Cancer Model for Experimental Cancer Therapy Studies

    Energy Technology Data Exchange (ETDEWEB)

    Ana J. Molinari; Romina F. Aromando; Maria E. Itoiz; Marcela A. Garabalino; Andrea Monti Hughes; Elisa M. Heber; Emiliano C. C. Pozzi; David W. Nigg; Veronica A. Trivillin; Amanda E. Schwint

    2012-07-01

    Normalization of tumor blood vessels improves drug and oxygen delivery to cancer cells. The aim of this study was to develop a technique to normalize blood vessels in the hamster cheek pouch model of oral cancer. Materials and Methods: Tumor-bearing hamsters were treated with thalidomide and were compared with controls. Results: Twenty eight hours after treatment with thalidomide, the blood vessels of premalignant tissue observable in vivo became narrower and less tortuous than those of controls; Evans Blue Dye extravasation in tumor was significantly reduced (indicating a reduction in aberrant tumor vascular hyperpermeability that compromises blood flow), and tumor blood vessel morphology in histological sections, labeled for Factor VIII, revealed a significant reduction in compressive forces. These findings indicated blood vessel normalization with a window of 48 h. Conclusion: The technique developed herein has rendered the hamster oral cancer model amenable to research, with the potential benefit of vascular normalization in head and neck cancer therapy.

  18. Vitamin K2 Ameliorates Damage of Blood Vessels by Glucocorticoid: a Potential Mechanism for Its Protective Effects in Glucocorticoid-induced Osteonecrosis of the Femoral Head in a Rat Model.

    Science.gov (United States)

    Zhang, Yuelei; Yin, Junhui; Ding, Hao; Zhang, Changqing; Gao, You-Shui

    2016-01-01

    Glucocorticoid has been reported to decrease blood vessel number and harm the blood supply in the femoral head, which is recognized to be an important mechanism of glucocorticoid-induced osteonecrosis of the femoral head (ONFH). To prevent glucocorticoid-induced ONFH, medication that promotes both bone formation and angiogenesis would be ideal. Vitamin K2 has been revealed to play an important role in bone metabolism; however, few studies have focused on the effect of Vitamin K2 on new vascular formation. Thus, this study aimed to investigate whether Vitamin K2 promoted new blood vessel formation in the presence of glucocorticoids, both in vitro and in vivo. The effect of Vitamin K2 on viability, migration, in vitro tube formation, and VEGF, vWF, CD31, KDR, Flt and PDGFB in EAhy926 incubated with or without dexamethasone were elucidated. VEGF, TGF-β and BMP-2, angiogenesis-related proteins secreted by osteoblasts, were also detected in the osteoblast-like cell line of MG63. In addition, blood vessels of the femoral head in rats administered with or without methylprednisolone and Vitamin K2 were evaluated using angiography and CD31 staining. In vitro studies showed that Vitamin K2 significantly protected endothelial cells from dexamethasone-induced apoptosis, promoted endothelial cell migration and in vitro tube formation. Angiogenesis-related proteins both in EAhy926 and MG63 were also upregulated by Vitamin K2 when cotreated with dexamethasone. In vivo studies showed enhanced blood vessel volume and CD31-positive staining cells in rats cotreated with VK2 and methylprednisolone compared to rats treated with methylprednisolone only. Collectively, Vitamin K2 has the ability to promote angiogenesis in vitro and to ameliorate vessels of the femoral head in glucocorticoid-treated rats in vivo, indicating that Vitamin K2 is a promising drug that may be used to prevent steroid-induced ONFH.

  19. Comparison of elastic--plastic and variable modulus-cracking constitutive models for prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Anderson, C.A.; Smith, P.D.

    1978-01-01

    The variable modulus-cracking model is capable of predicting the behavior of reinforced concrete structures (such as the reinforced plate under transverse pressure described previously) well into the range of nonlinear behavior including the prediction of the ultimate load. For unreinforced thick-walled concrete vessels under internal pressure the use of elastic--plastic concrete models in finite element codes enhances the apparent ductility of the vessels in contrast to variable modulus-cracking models that predict nearly instantaneous rupture whenever the tensile strength at the inner wall is exceeded. For unreinforced thick-walled end slabs representative of PCRV heads, the behavior predicted by finite element codes using variable modulus-cracking models is much stiffer in the nonlinear range than that observed experimentally. Although the shear type failures and crack patterns that are observed experimentally are predicted by such concrete models, the ultimate load carrying capacity and vessel-ductility are significantly underestimated. It appears that such models do not adequately model such features as aggregate interlock that could lead to an enhanced vessel reserve strength and ductility

  20. Advanced nondestructive examination of the reactor vessel head penetration tube welds

    International Nuclear Information System (INIS)

    Cvitanovic, M.; Zado, V.

    1996-01-01

    Beside a referent code examination requirements, appearance of the service induced flaws on the Reactor Vessel Head (RVH) penetration tube welds forced development of remotely operated examination tools and techniques. Several systems were developed for examination of RVH PWR type while only one system for examination of VVER - 440 type RVH has been developed by Inetec. In this article the most advanced RVH VVER - 440 type examination techniques such as ultrasonic, eddy current and visual testing techniques as well as remotely operated tool are described. (author)

  1. Stresses in transition region of VVER-1000 reactor vessels

    International Nuclear Information System (INIS)

    Namgung, I.; Nguye, T.L.

    2014-01-01

    Most of the western PWR reactor's bottom head is hemi-spherical shape, however for Russian designed VVER family of reactor it is ellipsoidal shape. The transition region from shell side to ellipsoidal head and transition top flange to cylindrical shell develop higher stress concentration than western PWR reactor vessel. This region can be modeled as conical shell with varying thickness. The theoretical derivation of stress in the thick-walled conical cylinder with varying thickness was developed and shown in detail. The results is applied to VVER-1000 reactor vessel of which shell to bottom ellipsoidal shell and shell to upper flange were investigated for stress field. The theoretical calculations were also compared with FEM solutions. An axisymmetric 3D model of VVER-1000 reactor vessel (without closure head) FEM model was created and internal hydrostatic pressure boundary condition was applied. The stress results from FEM and theoretical calculation were compared, and the discrepancies and accuracies of the theoretical results were discussed. (author)

  2. Stresses in transition region of VVER-1000 reactor vessels

    Energy Technology Data Exchange (ETDEWEB)

    Namgung, I. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Nguye, T.L. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); National Research Inst. of Mechanical Engineering, Hanoi City, Vietnam (China)

    2014-07-01

    Most of the western PWR reactor's bottom head is hemi-spherical shape, however for Russian designed VVER family of reactor it is ellipsoidal shape. The transition region from shell side to ellipsoidal head and transition top flange to cylindrical shell develop higher stress concentration than western PWR reactor vessel. This region can be modeled as conical shell with varying thickness. The theoretical derivation of stress in the thick-walled conical cylinder with varying thickness was developed and shown in detail. The results is applied to VVER-1000 reactor vessel of which shell to bottom ellipsoidal shell and shell to upper flange were investigated for stress field. The theoretical calculations were also compared with FEM solutions. An axisymmetric 3D model of VVER-1000 reactor vessel (without closure head) FEM model was created and internal hydrostatic pressure boundary condition was applied. The stress results from FEM and theoretical calculation were compared, and the discrepancies and accuracies of the theoretical results were discussed. (author)

  3. Assessment of reactor vessel integrity (ARVI)

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden)]. E-mail: sehgal@ne.kth.se; Karbojian, A. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Giri, A. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Kymaelaeinen, O. [FortumEngNP (Finland); Bonnet, J.M. [CEA (France); Ikkonen, K. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Sairanen, R. [VTT (Finland); Bhandari, S. [FRAMATOME (France); Buerger, M. [USTUTT (Germany); Dienstbier, J. [NRI Rez (Czech Republic); Techy, Z. [VEIKI (Hungary); Theofanous, T. [UCSB (United States)

    2005-02-01

    The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants. The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, which was concerned with the effects of stratification and metal layer on the thermal loads on the lower head wall during melt pool convection. The second experimental project was conducted at ULPU facility, which provided data and correlations of CHF due to the external cooling of the lower head.

  4. Different Recipient Vessels for Free Microsurgical Fibula Flaps in the Treatment of Avascular Necrosis of the Femoral Head: A Systematic Review and Meta-analysis.

    Science.gov (United States)

    Tu, Yiji; Chen, Zenggan; Lineaweaver, William Charles; Zhang, Feng

    2017-12-01

    Several recipient vessels can be used in free microsurgical fibula flaps (MFFs) for the treatment of avascular necrosis of the femoral head (ANFH). Few articles investigate the influence of different recipient vessels on outcomes of MFF for ANFH. A comprehensive literature search of databases including PubMed-Medline, Ovid-Embase, and Cochrane Library was performed to collect the related studies. The Medical Subject Headings used were "femur head necrosis" and "bone transplantation." The relevant words in title or abstract included but not limited to "fibula flap," "fibular flap," "vascularized fibula," "vascularized fibular," "free fibula," "free fibular," "femoral head necrosis," "avascular necrosis of femoral head," and "ischemic necrosis of femoral head." The methodological index for nonrandomized studies was adopted for assessing the studies included in this review. Finally, 15 studies encompassing a total of 1267 patients (1603 hips) with ANFH were pooled in the overall analysis. Recipient vessels for MFF included the ascending branch of the lateral circumflex femoral artery and vein in 8 studies, descending branch of the lateral circumflex femoral artery and vein in 2 studies, second perforating branch of the deep femoral artery and vein in 4 studies, and inferior gluteal artery and vein in 1 study. Preoperative and postoperative average Harris hip score and pooled analyses of the rate of conversion, radiographic progression, and hip surgery-related complications showed no significant difference on the outcomes of MFF on ANFH between using different recipient vessels. Different recipient vessels did not affect outcomes in MFF procedures for ANFH. High-quality randomized controlled trials and prospective studies would be necessary to clarify reliable advantages and disadvantages between different recipient vessels. Until then, surgeons are justified in using ascending branch of the lateral circumflex femoral artery and vein, descending branch of the lateral

  5. Containment vessel bottom head transport and lifting technique

    International Nuclear Information System (INIS)

    Zheng Donghong; Tian Shiyong; Hu Dequan; Xiao Hongtao

    2013-01-01

    The challengeable transport and lifting techniques and high safety assurance measures are needed for the onsite construction of the AP1000 containment vessel bottom head (CVBH), which is a large component with heavy weight, big size, high center of gravity, and easy to deformation. During transport, the infra structural road foundation is heavily loaded with big turning radius, and the requirement for synchronization of transport vehicles is strict. During lifting, the crane lifting capacities are high, requirement for the lifting and rigging tools is strict, nuclear island being put into place is difficult, and the crane operating foundation is heavily loaded. The transport and lifting techniques and safety assurance measures for CVBH are elaborated in detail, so as to provide a reference for the follow-up transport and lifting of large components of nuclear island. (authors)

  6. Proceedings of the Workshop on in-vessel core debris retention and coolability

    International Nuclear Information System (INIS)

    1999-01-01

    This conference on in-vessel core debris retention and coolability is composed of 37 papers grouped in three sessions: session 1 (Keynote papers: Key phenomena of late phase core melt progression, accident management strategies and status quo of severe fuel damage codes, In-vessel retention as a severe accident management scheme, GAREC analyses in support of in-vessel retention concept, Latest findings of RASPLAV project); session 2 - Experiments and model development with five sub-sessions: sub-session 1 (Debris bed heat transfer: Debris and Pool Formation/Heat Transfer in FARO-LWR: Experiments and Analyses, Evaporation and Flow of Coolant at the Bottom of a Particle-Bed modelling Relocated Debris, Investigations on the Coolability of Debris in the Lower Head with WABE-2D and MESOCO-2D, Uncertainty and Sensitivity Analysis of the Heat Transfer Mechanisms in the Lower Head, Simulation of the Arrival and Evolution of Debris in a PWR Lower Head with the SFD ICARE2 code), sub-session 2 (Corium properties, molten pool natural convection, and crust formation: Physico-chemistry and corium properties for in-vessel retention, Experimental data on heat flux distribution from volumetrically heated pool with frozen boundaries, Thermal hydraulic phenomena in corium pools - numerical simulation with TOLBIAC and experimental validation with BALI, TOLBIAC code simulations of some molten salt RASPLAV experiments, SIMECO experiments on in-vessel melt pool formation and heat transfer with and without a metallic layer, Numerical investigation of turbulent natural convection heat transfer in an internally-heated melt pool and metallic layer, Current status and validation of CON2D and 3D code, Free convection of heat-generating fluid in a constrained during experimental simulation of heat transfer in slice geometry), sub-session 3 (Gap formation and gap cooling: Quench of molten aluminum oxide associated with in-vessel debris retention by RPV internal water, Experimental investigations

  7. In-Vessel Retention Modeling Capabilities of SCDAP/RELAP5-3DC

    International Nuclear Information System (INIS)

    Knudson, D.L.; Rempe, J.L.

    2002-01-01

    Molten core materials may relocate to the lower head of a reactor vessel in the latter stages of a severe accident. Under such circumstances, in-vessel retention (IVR) of the molten materials is a vital step in mitigating potential severe accident consequences. Whether IVR occurs depends on the interactions of a number of complex processes including heat transfer inside the accumulated molten pool, heat transfer from the molten pool to the reactor vessel (and to overlying fluids), and heat transfer from exterior vessel surfaces. SCDAP/RELAP5-3D C has been developed at the Idaho National Engineering and Environmental Laboratory to facilitate simulation of the processes affecting the potential for IVR, as well as processes involved in a wide variety of other reactor transients. In this paper, current capabilities of SCDAP/RELAP5-3D C relative to IVR modeling are described and results from typical applications are provided. In addition, anticipated developments to enhance IVR simulation with SCDAP/RELAP5-3D C are outlined. (authors)

  8. Re-expression of pro-fibrotic, embryonic preserved mediators in irradiated arterial vessels of the head and neck region.

    Science.gov (United States)

    Möbius, Patrick; Preidl, Raimund H M; Weber, Manuel; Amann, Kerstin; Neukam, Friedrich W; Wehrhan, Falk

    2017-11-01

    Surgical treatment of head and neck malignancies frequently includes microvascular free tissue transfer. Preoperative radiotherapy increases postoperative fibrosis-related complications up to transplant loss. Fibrogenesis is associated with re-expression of embryonic preserved tissue developmental mediators: osteopontin (OPN), regulated by sex-determining region Y‑box 9 (Sox9), and homeobox A9 (HoxA9) play important roles in pathologic tissue remodeling and are upregulated in atherosclerotic vascular lesions; dickkopf-1 (DKK1) inhibits pro-fibrotic and atherogenic Wnt signaling. We evaluated the influence of irradiation on expression of these mediators in arteries of the head and neck region. DKK1, HoxA9, OPN, and Sox9 expression was examined immunohistochemically in 24 irradiated and 24 nonirradiated arteries of the lower head and neck region. The ratio of positive cells to total cell number (labeling index) in the investigated vessel walls was assessed semiquantitatively. DKK1 expression was significantly decreased, whereas HoxA9, OPN, and Sox9 expression were significantly increased in irradiated compared to nonirradiated arterial vessels. Preoperative radiotherapy induces re-expression of embryonic preserved mediators in arterial vessels and may thus contribute to enhanced activation of pro-fibrotic downstream signaling leading to media hypertrophy and intima degeneration comparable to fibrotic development steps in atherosclerosis. These histopathological changes may be promoted by HoxA9-, OPN-, and Sox9-related inflammation and vascular remodeling, supported by downregulation of anti-fibrotic DKK1. Future pharmaceutical strategies targeting these vessel alterations, e. g., bisphosphonates, might reduce postoperative complications in free tissue transfer.

  9. Ex-Vessel Core Melt Modeling Comparison between MELTSPREAD-CORQUENCH and MELCOR 2.1

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farmer, Mitchell [Argonne National Lab. (ANL), Argonne, IL (United States); Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-03-01

    System-level code analyses by both United States and international researchers predict major core melting, bottom head failure, and corium-concrete interaction for Fukushima Daiichi Unit 1 (1F1). Although system codes such as MELCOR and MAAP are capable of capturing a wide range of accident phenomena, they currently do not contain detailed models for evaluating some ex-vessel core melt behavior. However, specialized codes containing more detailed modeling are available for melt spreading such as MELTSPREAD as well as long-term molten corium-concrete interaction (MCCI) and debris coolability such as CORQUENCH. In a preceding study, Enhanced Ex-Vessel Analysis for Fukushima Daiichi Unit 1: Melt Spreading and Core-Concrete Interaction Analyses with MELTSPREAD and CORQUENCH, the MELTSPREAD-CORQUENCH codes predicted the 1F1 core melt readily cooled in contrast to predictions by MELCOR. The user community has taken notice and is in the process of updating their systems codes; specifically MAAP and MELCOR, to improve and reduce conservatism in their ex-vessel core melt models. This report investigates why the MELCOR v2.1 code, compared to the MELTSPREAD and CORQUENCH 3.03 codes, yield differing predictions of ex-vessel melt progression. To accomplish this, the differences in the treatment of the ex-vessel melt with respect to melt spreading and long-term coolability are examined. The differences in modeling approaches are summarized, and a comparison of example code predictions is provided.

  10. Evaluation of Thermal Load to the Lower Head Vessel Using the ASTEC Computer Code

    International Nuclear Information System (INIS)

    Park, Raejoon; Ahn, Kwangil

    2013-01-01

    The thermal load from the corium to the lower head vessel in the APR (Advanced Power reactor) 1400 during a small break loss of coolant accident (SBLOCA) without a safety injection (SI) has been evaluated using the ASTEC (Accident Source Term Evaluation Code) computer code, which has been developed as a part of the EU (European Union)-SARNET (Severe Accident Research NET work) program. The ASTEC results predict that the reactor vessel did not fail by using an ERVC, in spite of the large melting of the reactor vessel wall in a two-layer formation case of the SBLOCA in the APR1400. The outer surface conditions of the temperature and heat transfer coefficient are not effective on the vessel geometry change, which are preliminary results. A more detailed analysis of the main parameter effects on the corium behavior in the lower plenum is necessary to evaluate the IVR-ERVC in the APR1400, in particular, for a three-layer formation of the TLFW. Comparisons of the present results with others are necessary to verify and apply them to the actual IVR-ERVC evaluation in the APR1400

  11. Debris interactions in reactor vessel lower plena during a severe accident. II. Integral analysis

    International Nuclear Information System (INIS)

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

    1996-01-01

    For pt.I see ibid., p.147-63, 1996. The integral physico-numerical model for the reactor vessel lower head response has been exercised for the TMI-2 accident and possible severe accident scenarios in PWR and BWR designs. The proposed inherent cooling mechanism of the reactor material creep and subsequent water ingression implemented in this predictive model provides a consistent representation of how the debris was finally cooled in the TMI-2 accident and how the reactor lower head integrity was maintained during the course of the incident. It should be recalled that in order for this strain to occur, the vessel lower head had to achieve temperatures in excess of 1000 C. This is certainly in agreement with the temperatures determined by metallographic examinations during the TMI-2 vessel inspection program. The integral model was also applied to typical PWR and BWR lower plena with and without structures under pressurized conditions spanning the first relocation of core material to the reactor vessel failure due to creep without recovery actions. The design application results are presented with particular attention being focused on water ingression into the debris bed through the gap formed between the debris and the vessel wall. As an illustration of the accident management application, the lower plenum with structures was recovered after an extensive amount of creep had damaged the vessel wall. The computed lower head temperatures were found to be significantly lower (by more than 300 K in this particular example) with recovery relative to the case without recovery. This clearly demonstrates the potential for in-vessel cooling of the reactor vessel without a need to externally submerge the lower head should such a severe accident occur as core melting and relocation. (orig.)

  12. Evaluation of J-groove weld residual stress and crack growth rate of PWSCC in reactor pressure vessel closure head

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Seung Hyuk; Ryu, Tae Young; Park, Seung Hyun; Won, Min Gu; Kang, Seok Jun; Kim, Moon Ki; Choi, Jae Boong [Sungkyunkwan University, Suwon (Korea, Republic of); Lee, Kyoung Soo; Lee, Sung Ho [Korea Hydro and Nuclear Power, Daejeon (Korea, Republic of)

    2015-03-15

    Over the last decade, primary water stress corrosion cracking (PWSCC) has been frequently found in pressurized water reactor (PWR) applications. Especially, PWSCC has occurred in long-term operated PWRs. As this phenomenon leads to serious accidents, we must be beforehand with the anticipated problems. A typical PWR consists of J-groove welded components such as reactor pressure vessel closure head and nozzles. Reactor pressure vessel closure head is made of SA508 and it is covered by cladding. Alloy 600 is used for nozzles. And J-groove weld is conducted with alloy 82/182. Different material properties of these metals lead to residual stress and PWSCC consequentially. In this study, J-groove weld residual stress was investigated by a three-dimensional finite element analysis with an actual asymmetric J-groove weld model and process of construction. Also crack growth rate of PWSCC was evaluated from cracks applied on the penetration nozzles. Based on these two values, one cannot only improve the structural integrity of PWR, but also explain PWSCC behavior such that high residual stress at the J-groove weld area causes crack initiation and propagation through the surface of nozzles. In addition, crack behavior was predicted at the various points around the nozzle.

  13. Evaluation of J-groove weld residual stress and crack growth rate of PWSCC in reactor pressure vessel closure head

    International Nuclear Information System (INIS)

    Oh, Seung Hyuk; Ryu, Tae Young; Park, Seung Hyun; Won, Min Gu; Kang, Seok Jun; Kim, Moon Ki; Choi, Jae Boong; Lee, Kyoung Soo; Lee, Sung Ho

    2015-01-01

    Over the last decade, primary water stress corrosion cracking (PWSCC) has been frequently found in pressurized water reactor (PWR) applications. Especially, PWSCC has occurred in long-term operated PWRs. As this phenomenon leads to serious accidents, we must be beforehand with the anticipated problems. A typical PWR consists of J-groove welded components such as reactor pressure vessel closure head and nozzles. Reactor pressure vessel closure head is made of SA508 and it is covered by cladding. Alloy 600 is used for nozzles. And J-groove weld is conducted with alloy 82/182. Different material properties of these metals lead to residual stress and PWSCC consequentially. In this study, J-groove weld residual stress was investigated by a three-dimensional finite element analysis with an actual asymmetric J-groove weld model and process of construction. Also crack growth rate of PWSCC was evaluated from cracks applied on the penetration nozzles. Based on these two values, one cannot only improve the structural integrity of PWR, but also explain PWSCC behavior such that high residual stress at the J-groove weld area causes crack initiation and propagation through the surface of nozzles. In addition, crack behavior was predicted at the various points around the nozzle.

  14. Structural criteria for extreme dynamic internal pressure loadings of vessels and closure heads

    International Nuclear Information System (INIS)

    Bitner, J.L.

    1985-01-01

    The criteria protect against tensile plastic instability and local ductile rupture failure modes. To minimize the number of critical areas that may need more rigorous analytical methods, a screening criterion for limiting the membrane, bending and local stresses is defined. The stresses for this criterion are calculated from either simple and economical elastic dynamic or equivalent static methods. For the critical areas that remain, a strain-based criterion for strains derived from dynamic, inelastic methods is given. To assure that the criteria are properly applied, guidelines are outlined for controlling methods for deriving stresses and strains, for selecting appropriate material properties and for addressing specific dominating parameters that affect the validity of the analysis. The application of the criteria to a complex liquid metal fast breeder reactor vessel and closure head and the subsequent experimental verification of the results by several scale model experiments are summarized. (orig./HP)

  15. In-vessel inspection before head removal: TMI II, Phase III (tooling and systems design and verification)

    International Nuclear Information System (INIS)

    Carter, G.S.; Ryan, R.F.; Pieleck, A.W.; Bibb, H.Q.

    1982-09-01

    Under EG and G contract K-9003 to General Public Utilities Corporation, a Task Order was assigned to Babcock and Wilcox to develop and provide equipment to facilitate early assessment of core damage in the Three Mile Island Unit 2 reactor vessel head. Described is the work performed, the equipment developed, and the tests conducted with this equipment on various mockups used to simulate the constraints inside and outside the reactor vessel that affect the performance of the inspection. The tooling developed provides several methods of removing a few control rod drive leadscrews from the reactor, thereby providing paths into which cameras and lights may be inserted to permit video viewing of many potentially damaged areas in the reactor vessel. The tools, equipment, and cameras demonstrated that these tasks could be accomplished

  16. Persistent trigeminal artery/persistent trigeminal artery variant and coexisting variants of the head and neck vessels diagnosed using 3 T MRA

    International Nuclear Information System (INIS)

    Bai, M.; Guo, Q.; Li, S.

    2013-01-01

    Aim: To report the prevalence and characteristic features of persistent trigeminal artery (PTA), PTA variant (PTAV), and other variants of the head and neck vessels, identified using magnetic resonance angiography (MRA). Materials and methods: The three-dimensional (3D) time of flight (TOF) MRA and 3D contrast-enhanced (CE) MRA images of 6095 consecutive patients who underwent 3 T MRA at Liaocheng People's Hospital from 1 September 2008 through 31 May 2012 were retrospectively reviewed and analysed. Thirty-two patients were excluded because of suboptimal image quality or internal carotid artery (ICA) occlusion. Results: The prevalence of both PTA and PTAV was 0.63% (PTA, 26 cases; PTAV, 12 cases). The prevalence of coexisting variants of the head and neck vessels in cases of PTA/PTAV was 52.6% (20 of 38 cases). The vascular variants that coexisted with cases of PTA/PTAV were as follows: the intracranial arteries varied in 10 cases, the origin of the supra-aortic arteries varied in nine cases, the vertebral artery (VA) varied in 14 cases, and six cases displayed fenestrations. Fifteen of the 20 cases contained more than two types of variants. Conclusion: The prevalence of both PTA and PTAV was 0.63%. Although PTA and PTAV are rare vascular variants, they frequently coexist with other variants of the head and neck vessels. Multiple vascular variations can coexist in a single patient. Recognizing PTA, PTAV, and other variants of the head and neck vessels is crucial when planning a neuroradiological intervention or surgery. Recognizing the medial PTA is very important in clinical practice when performing trans-sphenoidal surgery on the pituitary as failure to do so could result in massive haemorrhage

  17. Investigation of a weld defect, reactor vessel head Ringhals 2

    International Nuclear Information System (INIS)

    Embring, G.; Pers-Anderson, E.B.

    1994-01-01

    During the summer-outage 1993 Ringhals unit 2 vessel head was inspected at weld-area of Alloy 182. One major defect was found Two plus two ''boat-samples'' were taken out from the zone between the weld and the stainless cladding. All samples were sent to Studsviks laboratories for detailed investigations. The metallographic and fractographic investigations revealed that the major weld-defect had been there from manufacturing. The defect was located between the Alloy 182-buttering and the pressure vessel steel SA 533 grB cl 1. No indications of PWSCC or IDSCC were found. An inspection programme was defined. Different types of reference blocks were provided by Ringhals in cooperation with ABB TRC. Reference reflectors of type flat bottom hole (FBH) and eroded notches (EDM), with different sizes and separation were manufactured. One weld sample with manufacturing defects -lack of fusion and slag was inclusions- was present. ABB TRC performed UT inspection in the gap between the penetration and the thermal sleeve. Inspection results like defect identification, defect separation and defect sizing accuracy were compared with result from the destructive inspection. No relevant additional defects were found. An analysing and repair program was performed. A special designed disc sealed off the defect area. (authors). 5 figs., 3 refs

  18. Experiences concerning reactor pressure vessel head penetration inspections; Erfahrungen mit Pruefungen von Reaktordruckbehaelter-Deckeldurchfuehrungen

    Energy Technology Data Exchange (ETDEWEB)

    Debnar, Angelika [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2009-07-01

    Globally observed damage at the control rod drive mechanism nozzles in PWR-type reactors (Bugey-3, Oconee 1,2,3 and ANO-1, David Besse) have triggered enhanced inspection of reactor pressure vessel (RPV) head penetrations. In Germany the regulations require a periodic inspection especially of dissimilar welds. Westinghouse has developed an automated measuring system for RPV heads aimed to inspect welded joints at open nozzles of nozzles with thermosleeves. The testing technology with remote controlled robotics is supposed to perform a weld inspection as complete as possible, restraints result from constructive difficulties for the accessibility. The new gap-scanner DE2008 was qualified at the mock-up and was implemented into the periodic in-service inspection of Neckarwestheim-1.

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  20. The modeling and analysis of in-vessel corium/structure interaction in boiling water reactors

    International Nuclear Information System (INIS)

    Podowski, M.Z.; Kurul, N.; Kim, S.-W.; Baltyn, W.; Frid, W.

    1997-01-01

    A complete stand-alone state-of-the-art model has been developed of the interaction between corium debris in the lower plenum and the RPV walls and internal structures, including the vessel failure mechanisms. This new model has been formulated as a set of consistent computer modules which could be linked with other existing models and/or computer codes. The combined lower head and lower plenum modules were parametrically tested and applied to predict the consequences of a hypothetical station blackout in a Swedish BWR. (author)

  1. An analysis of critical heat flux on the external surface of the reactor vessel lower head

    International Nuclear Information System (INIS)

    Yang, Soo Hyung; Baek, Won Pil; Chang, Soon Heung

    1999-01-01

    CHF (Critical heat flux) on the external surface of the reactor vessel lower head is major key in the evaluation on the feasibility of IVR-EVC (In-Vessel Retention through External Vessel Cooling) concept. To identify the CHF on the external surface, considerable works have been performed. Through the review on the previous works related to the CHF on the external surface, liquid subcooling, induced flow along the external surface, ICI (In-Core Instrument) nozzle and minimum gap are identified as major parameters. According to the present analysis, the effects of the ICI nozzle and minimum gap on CHF are pronounced at the upstream of test vessel: on the other hand, the induced flow considerably affects the CHF at downstream of test vessel. In addition, the subcooling effect is shown at all of test vessel, and decreases with the increase in the elevation of test vessel. In the real application of the IVR-EVC concept, vertical position is known as a limiting position, at which thermal margin is the minimum. So, it is very important to precisely predict the CHF at vertical position in a viewpoint of gaining more thermal margins. However, the effects of the liquid subcooling and induced flow do not seem to be adequately included in the CHF correlations suggested by previous works, especially at the downstream positions

  2. 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)

  3. Advanced in-vessel retention design for next generation risk management

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Y; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    1998-12-31

    In the TMI-2 accident, approximately twenty (20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However,one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 deg C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant (KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. 15 refs., 5 figs., 1 tab. (Author)

  4. Advanced in-vessel retention design for next generation risk management

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Y.; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    1997-12-31

    In the TMI-2 accident, approximately twenty (20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However,one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 deg C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant (KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. 15 refs., 5 figs., 1 tab. (Author)

  5. A Study on the Coupled FEM-Analysis for Reactor Vessel Lower Head of APR1400 under the Severe Accident Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyonam; Namgung, Ihn [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-10-15

    For the stabilization of the RPV the in-vessel retention strategy with external reactor vessel cooling (IVR-ERVC) is adopted in APR1400. Under this severe accident condition, a good understanding of the mechanical behavior of the reactor vessel lower head (RVLH) is necessary both for verification of structural integrity and for improving the design applying appropriate accident mitigation strategies. The purpose of this study is to develop the analysis method of the RVLH with thermo-mechanical analysis using FEM tool (ANSYS v.15) in case of core-melting severe accident condition, and then analyze the RVLH of APR1400 including creep behavior. The plastic strain can be the major cause of lower head failure on the reactor vessel, and the creep cannot be not negligible factor of the failure under the severe accident condition. In the study, we applied constant convection coefficient at assumed temperature on the outside wall of RPV and substitute creep data of SA-508. In addition, it was found that the steel ablation at the interface between corium and vessel steel is not only a thermal phenomenon in the METCOR experiments. Corrosion processes and the formation of eutectics lead to the erosion of the vessel steel at temperatures that are significantly lower than the melting temperature of steel. It called thermo-chemical attack of the corium (corrosion). Reduced wall thickness because of the thermo-chemical effect by corium increase the equivalent plastic strain, and decrease the minimum time to reach 20% creep strain.

  6. 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

  7. 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

  8. Review of the TMI-2 accident evaluation and vessel investigation projects

    Energy Technology Data Exchange (ETDEWEB)

    Ladekarl Thomsen, Knud

    1998-03-01

    The results of the TMI-2 Accident Evaluation Programme and the Vessel Investigation Project have been reviewed as part of a literature study on core meltdown and in-vessel coolability. The emphasis is placed on the late phase melt progression, which is of special relevance to the NKS-sponsored RAK-2.1 project on Severe Accident Phenomenology. The body of the report comprises three main sections, The TMI-2 Accident Scenario, Core Region and Relocation Path Investigations, and Lower Head Investigations. In the final discussion, the lower head gap formation mechanism is explained in terms of thermal contraction and fracturing of the debris crust. This model seems more plausible than the MAAP model based on creep expansion of the lower head. (au) 1 tab., 33 ills., 31 refs.

  9. TMI-2 Vessel Investigation Project Metallurgical Program

    International Nuclear Information System (INIS)

    Diercks, D.R.; Neimark, L.A.

    1990-01-01

    The TMI-2 [Three Mile Island unit 2] Vessel Investigation Project Metallurgical Program at Argonne National Laboratory is a part of the international TMI-2 Vessel Investigation Project being conducted jointly by the U.S. Nuclear Regulatory Commission and the Organization for Economic Co-operation and Development (OECD). The overall project consists of three phases, namely (1) recovery of material samples from the lower head of the TMI-2 reactor, (2) examination and analysis of the lower head samples and the preparation and testing of archive material subjected to a similar thermal history, and (3) procurement, examination, and analysis of companion core material located adjacent to or near the lower head material. The specific objectives of the ANL Metallurgical Program, which accounts for a major portion of Phase 2, are to prepare metallographic and mechanical test specimen blanks from the TMI-2 lower head material, prepare similar test specimen blanks from suitable archive material subjected to the appropriate thermal processing, determine the mechanical properties of the lower vessel head and archive materials under the conditions of the core-melt accident, and assess the lower head integrity and margin-to-failure during the accident. The ANL work consists of three tasks: (1) archive materials program, (2) fabrication of metallurgical and mechanical test specimens from the TMI-2 pressure vessel samples, and (3) mechanical property characterization of TMI-2 lower pressure vessel head and archive material

  10. Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae

    2016-01-01

    If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study

  11. Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae [KEPCO Engineering and Construction Co. Ltd., Deajeon (Korea, Republic of)

    2016-10-15

    If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study.

  12. TMI-2 Vessel Investigation Project (VIP) Metallurgical Program

    International Nuclear Information System (INIS)

    Diercks, D.R.; Neimark, L.A.

    1990-06-01

    The TMI-2 Vessel Investigation Project (VIP) Metallurgical Program is a part of the international TMI-2 Vessel Investigation Project being conducting jointly by the US Nuclear Regulatory Commission and the Organization for Economic Co-operation and Development (OECD). The overall project consists of three phases, namely (1) recovery of material samples from the lower head of the TMI-2 reactor, (2) examination and analysis of the lower head samples and the preparation and testing of archive material subjected to a similar thermal history, and (3) procurement, examination, and analysis of companion core material located adjacent to or near the lower head material. The specific objectives of the ANL Metallurgical Program, which comprises a major portion of Phase 2, are to prepare metallographic and mechanical test specimen blanks from the TMI-2 lower head material, prepare similar test specimen blanks from suitable archive material subjected to the appropriate thermal processing, determine the mechanical properties of the lower vessel head and archive materials under the conditions of the core-melt accident, and assess the lower head integrity and margin-to-failure during the accident. The ANL work consists of three tasks: (1) archive materials program, (2) fabrication of metallurgical and mechanical test specimens from the TMI-2 pressure vessel samples, and (3) mechanical property characterization of TMI-2 lower pressure vessel head and archive material

  13. Effect on temperature of output of the core of the size of the break in the Upper Head of the vessel using TRACE5; Efecto sobre la Temperatura de Salida del Nucleo del Tamano de la Rotura en el Upper Head de la Vasija utilizando TRACE5

    Energy Technology Data Exchange (ETDEWEB)

    Querol, A.; Gallardo, S.; Verdu, G.

    2013-07-01

    Most (PWR) pressurized water reactors have thermocouples to detect overheating of the core since they are used to measure the temperature of exit of the nucleus (CET). However, it was found that in a small break (SBLOCA) located in the upper head of the vessel there is a delay between the measure of thermocouples and overheating of the core. This work is based on the simulation, using the code Thermo-hydraulic TRACE5, of the Test 6 - 1 the OECD/NEA rose project carried out in the experimental facility LSTF (Large Scale Test Facility). There have been different analyses in which geometric variables that can influence the model such as the size and location of the break, possible flow towards the break and the nodalization of the upper head of the vessel have been studied.

  14. Lining up device for the internal structures of a nuclear reactor vessel

    International Nuclear Information System (INIS)

    Silverblatt, B.L.

    1977-01-01

    The invention concerns a nuclear reactor of the type with a vessel, a vessel head carried at the top of this vessel by a core cylinder comprising a flange internally supported by the vessel, and an upper support structure supported between the core cylinder flange and the vessel head to align laterally the head, vessel, flange and support structure. A bottom key device is provided for lining up the flange, support structure and vessel, and an upper key device for laterally lining up support structure and the vessel head and for maintaining this alignment when they are removed simultaneously from the core cylinder and vessel. When re-assembling the reactor, the top support structure and the vessel head are lowered simultaneously so that an opening in the top alignment structure engages in the upper extension of the bottom alignment structure. A plurality of alignment stuctures may be utilised round the circumference of the reactor vessel. The disposition of the invention also facilitates the removal of the core cylinder from the reactor vessel. In this way, the alignment on re-assembly is ensured by the re-entry of the bottom extension under the flange of the core cylinder with the groove or keyway of the reactor vessel [fr

  15. Handling and carrying head for nuclear fuel assemblies and installation including this head

    International Nuclear Information System (INIS)

    Artaud, R.; Cransac, J.P.; Jogand, P.

    1986-01-01

    The present invention proposes a handling and carrying head ensuring efficiently the cooling of the nuclear fuel asemblies it transports so that any storage in liquid metal in a drum within or adjacent the reactor vessel is suppressed. The invention claims also a nuclear fuel handling installation including the head; it allows a longer time between loading and unloading campaigns and the space surrounding the reactor vessel keeps free without occupying a storage zone within the vessel [fr

  16. In-vessel core debris retention experiments. Final report

    International Nuclear Information System (INIS)

    1998-10-01

    The in-vessel cooling experimental program (Phase 1 and 2) was motivated by the survivability of the TMI lower vessel head during the TMI-2 accident. During that accident, molten debris relocation into the water filled lower head resulted in a localized hot spot in the lower head, but no lower head failure occurred. A postulated set of mechanisms which could be involved in and responsible for the survivability of the TMI lower head were identified and experimentally investigated as part of this program. These mechanisms included: the formation of a gap (contact resistance) between the relocated and frozen debris and the vessel wall was a key aspect of the in-vessel cooling mechanism; wall heatup due to the relocated debris in the presence of wall stress due to a pressure gradient across the vessel wall; gap growth due to a lack of debris adherence to the vessel wall and material creep of the heated vessel wall; and the potential for enhanced wall cooling due to gap growth. Each of these postulated mechanisms was investigated in this experimental program. This report summarizes the several insights and conclusions that were obtained from this experimental program. This report documents the entire set of five experiments completed in Phase 2 of this experimental program. Results from the Phase 1 effort were used to plan and select the Phase 2 test matrix. Conclusions from the Phase 1 and 2 experiments are identified and recommendations for future work are provided

  17. Bottom head failure program plan

    International Nuclear Information System (INIS)

    Meyer, R.O.

    1989-01-01

    Earlier this year the NRC staff presented a Revised Severe Accident Research Program Plan (SECY-89-123) to the Commission and initiated work on that plan. Two of the near-term issues in that plan involve failure of the bottom head of the reactor pressure vessel. These two issues are (1) depressurization and DCH and (2) BWR Mark I Containment Shell Meltthrough. ORNL has developed models for several competing failure mechanisms for BWRs. INEL has performed analytical and experimental work directly related to bottom head failure in connection with several programs. SNL has conducted a number of analyses and experimental activities to examine the failure of LWR vessels. In addition to the government-sponsored work mentioned above, EPRI and FAI performed studies on vessel failure for the Industry Degraded Core Rulemaking Program (IDCOR). EPRI examined the failure of a PWR vessel bottom head without penetrations, as found in some Combustion Engineering reactors. To give more attention to this subject as called for by the revised Severe Accident Research Plan, two things are being done. First, work previously done is being reviewed carefully to develop an overall picture and to determine the reliability of assumptions used in those studies. Second, new work is being planned for FY90 to try to complete a reasonable understanding of the failure process. The review and planning are being done in close cooperation with the ACRS. Results of this exercise will be presented in this paper

  18. Light water reactor lower head failure analysis

    International Nuclear Information System (INIS)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L.

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response

  19. Light water reactor lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

  20. Stress analysis in a non axisymmetric loaded reactor pressure vessel

    International Nuclear Information System (INIS)

    Albuquerque, Levi Barcelos; Assis, Gracia Menezes V. de; Miranda, Carlos Alexandre J.; Cruz, Julio Ricardo B.; Mattar Neto, Miguel

    1995-01-01

    In this work we intend to present the stress analysis of a PWR vessel under postulated concentrated loads. The vessel was modeled with Axisymmetric solid 4 nodes harmonic finite elements with the use of the ANSYS program, version 5.0. The bolts connecting the vessel flanges were modeled with beam elements. Some considerations were made to model the contact between the flanges. The perforated part of the vessel tori spherical head was modeled (with reduced properties due to its holes) to introduce its stiffness and loads but was not within the scope of this work. The loading consists of some usual ones, as pressure, dead weight, bolts preload, seismic load and some postulated ones as concentrated loads, over the vessel, modeled by Fourier Series. The results in the axisymmetric model are taken in terms of linearized stresses, obtained in some circumferential positions and for each position, in some sections along the vessel. Using the ASME Code (Section III, Division 1, Sub-section NB) the stresses are within the allowable limits. In order to draw some conclusions about stress linearization, the membrane plus bending stresses (Pl + Pb) are obtained and compared in some sections, using three different methods. (author)

  1. Computing the partial volume of pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Wiencke, Bent [Nestle USA, Corporate Engineering, 800 N. Brand Blvd, Glendale, CA 91203 (United States)

    2010-06-15

    The computation of the partial and total volume of pressure vessels with various type of head profiles requires detailed knowledge of the head profile geometry. Depending on the type of head profile the derivation of the equations can become very complex and the calculation process cumbersome. Certain head profiles require numerical methods to obtain the partial volume, which for most application is beyond the scope of practicability. This paper suggests a unique method that simplifies the calculation procedure for the various types of head profiles by using one common set of equations without the need for numerical or complex computation methods. For ease of use, all equations presented in this paper are summarized in a single table format for horizontal and vertical vessels. (author)

  2. 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.)

  3. Nuclear reactor vessel inspection apparatus

    International Nuclear Information System (INIS)

    Blackstone, E.G.; Lofy, R.A.; Williams, L.P.

    1979-01-01

    Apparatus for the in situ inspection of a nuclear reactor vessel to detect the location and character of flaws in the walls of the vessel, in the welds joining the various sections of the vessel, in the welds joining attachments such as nozzles, elbows and the like to the reactor vessel and in such attachments wherein an inspection head carrying one or more ultrasonic transducers follows predetermined paths in scanning the various reactor sections, welds and attachments

  4. Method and system for installing a layered vessel on location

    International Nuclear Information System (INIS)

    Pechacek, R.E.; Clay, E.J.

    1982-01-01

    A method and system for installing a layered vessel wherein the method includes the steps of constructing the bottom vessel head section in an inverted position mounting the bottom head section on the vessel foundation, erecting a generally cylindrical construction frame having a plurality of annular work stations; substantially simultaneously with the erection of the cylindrical construction frame, constructing onto the bottom head a cylindrical inside shell liner and a hemispherical upper head inside liner and adding layers to the inside shell from the bottom upwardly with the addition of such layers occurring substantially simultaneously at various of the annular work stations. A system for accomplishing these steps is provided, including particular method for constructing the bottom head, and further, an annularly movable crane assembly is provided for the work stations. (author)

  5. 3 D flow computations under a reactor vessel closure head

    International Nuclear Information System (INIS)

    Daubert, O.; Bonnin, O.; Hofmann, F.; Hecker, M.

    1995-12-01

    The flow under a vessel cover of a pressurised water reactor is investigated by using several computations and a physical model. The case presented here is turbulent, isothermal and incompressible. Computations are made with N3S code using a k-epsilon model. Comparisons between numerical and experimental results are on the whole satisfying. Some local improvements are expected either with more sophisticated turbulence models or with mesh refinements automatically computed by using the adaptive meshing technique which has been just implemented in N3S for 3D cases. (authors). 6 refs., 7 figs

  6. PDX vacuum vessel stress analysis

    International Nuclear Information System (INIS)

    Nikodem, Z.D.

    1975-01-01

    A stress analysis of PDX vacuum vessel is described and the summary of results is presented. The vacuum vessel is treated as a toroidal shell of revolution subjected to an internal vacuum. The critical buckling pressure is calculated. The effects of the geometrical discontinuity at the juncture of toroidal shell head and cylindrical outside wall, and the concavity of the cylindrical wall are examined. An effect of the poloidal field coil supports and the vessel outside supports on the stress distribution in the vacuum vessel is determined. A method evaluating the influence of circular ports in the vessel wall on the stress level in the vessel is outlined

  7. Heat transfer between relocated materials and the RPV lower head

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Kohriyama, T.

    2001-01-01

    Questions about the coolability of a continuous mass of relocated corium were raised during the Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) Post-accident examinations indicate that nearly half of the material that relocated to the vessel lower head during the TMI-2 accident formed a cohesive or ''continuous'' layer. TMI-2 VIP results and other evidence suggest that conduction through this continuous layer of solidified corium materials was assisted by other cooling mechanisms. Because increased knowledge about in-vessel coolability of corium materials may assist reactor designers in demonstrating that their concepts are passively safe, there is international interest in this topic. However, data are needed to identify what cooling mechanism(s) occurred and to develop a validated model for predicting this cooling. Corium cooling models significantly impact predictions for subsequent accident progression, such as the estimated time and mode of vessel failure. Hence, improved cooling models will provide a much needed, missing component of severe accident analyses. This paper provides a critical review of research investigating the coolability of corium relocating to a water-filled lower head. Where possible, existing models and data for predicting cooling are quantitatively compared; and governing relationships are identified. Key phenomena that should be incorporated into models for predicting this heat transfer are discussed, and deficiencies in current models and available data for predicting cooling are noted. Recommendations for improving these models and for obtaining data to validate these models are also provided. (author)

  8. State of the Art Report for the In-Vessel Late Core Melt Progression

    International Nuclear Information System (INIS)

    Kim, Hee Dong; Kang, Kyoung Ho; Park, Rae Joon

    2009-04-01

    The formation of corium pool in the reactor vessel lower head and its behavior is still an important issue. This issue is closely related to understanding of the core melting, its course, critical phases and timing during severe accidents and the influence of these processes on the accident progression, especially the evaluation of in-vessel retention by external reactor vessel cooling (IVR-ERVC) as a severe accident management strategy. The previous researches focused on the quisi-steady state behavior of molten corium pool in the lower head and related in-vessel retention problem. However, questions of the feasibility of the in-vessel retention concept for high power density reactor and uncertainties due to layering effect require further studies. These researches are rather essential to consider the whole evolution of the accident including formation and growth of the molten pool and the characteristic of corium arrival in the lower head and molten pool behavior after the core debris remelting. The general objective of the LIVE program performed at FzK is to study the corium pool formation and behavior with emphasis on the transient behavior through the large scale 3-D experiments. In this report, description of LIVE experimental facility and results of performance test are briefly summarized and the process to select the simulant is depicted. Also, the results of LIVE L1 and L2 tests and analytical models are included. These experimental results are very useful to development and verification of the model of molten corium pool behavior

  9. 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

  10. Three Mile Island unit 2 vessel investigation project. Conclusions and significance

    International Nuclear Information System (INIS)

    Beckjord, E.S.

    1994-01-01

    At the conclusion of the TMI-2 Vessel Investigation Project, additional insights about the accident have been gained, specifically in the area of reactor vessel integrity and the conditions of the lower head of the reactor vessel. This paper discusses three topics: the evolving views about the TMI-2 accident scenario over time, the technical conclusions of the TMI-2 VIP (recovery of samples from the vessel lower head), and the broad significance of these findings (accident management). 4 refs

  11. Histomorphological changes of vessel structure in head and neck vessels following preoperative or postoperative radiotherapy

    International Nuclear Information System (INIS)

    Schultze-Mosgau, S.; Wehrhan, F.; Wiltfang, J.; Grabenbauer, G.G.; Sauer, R.; Roedel, F.; Radespiel-Troeger, M.

    2002-01-01

    Patients and Methods: In 348 patients (October 1995-March 2002) receiving primarly or secondarily 356 microvascular hard- and soft tissue reconstruction, a total of 209 vessels were obtained from neck recipient vessels and transplant vessels during anastomosis. Three groups were analysed: group 1 (27 patients) treated with no radiotherapy or chemotherapy; group 2 (29 patients) treated with preoperative irradiation (40-50 Gy) and chemotherapy (800 mg/m 2 /day 5-FU and 20 mg/m 2 /day cisplatin) 1.5 months prior to surgery; group 3 (20 patients) treated with radiotherapy (60-70 Gy) (median interval 78.7 months; IQR: 31.3 months) prior to surgery. From each of the 209 vessel specimens, 3 sections were investigated histomorphometrically, qualitatively and quantitatively (ratio media area/total vessel area) by NIH-Image-digitized measurements. To evaluate these changes as a function of age, radiation dose and chemotherapy, a statistical analysis was performed using an analysis of covariance and χ 2 tests (p > 0.05, SPSS V10). Results: In group 3, qualitative changes (intima dehiscence, hyalinosis) were found in recipient arteries significantly more frequently than in groups 1 and 2. For group 3 recipient arteries, histomorphometry revealed a significant decrease in the ratio media area/total vessel area (median 0.51, IQR 0.10) in comparison with groups 1 (p = 0.02) (median 0.61, IQR 0.29) and 2 (p = 0.046) (median 0.58, IQR 0.19). No significant difference was found between the vessels of groups 1 and 2 (p = 0.48). There were no significant differences in transplant arteries and recipient or transplant veins between the groups. Age and chemotherapy did not appear to have a significant influence on vessel changes in this study (p > 0.05). Conclusions: Following irradiation with 60-70 Gy, significant qualitative and quantitative histological changes to the recipient arteries, but not to the recipient veins, could be observed. In contrast, irradiation at a dose of 40-50 Gy

  12. Nuclear reactor support and seismic restraint with in-vessel core retention cooling features

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, Tyler A.; Edwards, Michael J.

    2018-01-23

    A nuclear reactor including a lateral seismic restraint with a vertically oriented pin attached to the lower vessel head and a mating pin socket attached to the floor. Thermally insulating materials are disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head.

  13. Simulation of time of flight defraction signals for reactor vessel head penetrations

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Tae Hun; Kim, Young Sik; Lee, Jeong Seok [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2016-08-15

    The simulation of nondestructive testing has been used in the prediction of the signal characteristics of various defects and in the development of the procedures. CIVA, a simulation tool dedicated to nondestructive testing, has good accuracy and speed, and provides a three-dimensional graphical user interface for improved visualization and familiar data displays consistent with an NDE technique. Even though internal validations have been performed by the CIVA software development specialists, an independent validation study is necessary for the assessment of the accuracy of the software prior to practical use. In this study, time of flight diffraction signals of ultrasonic inspection of a calibration block for reactor vessel head penetrations were simulated using CIVA. The results were compared to the experimentally inspected signals. The accuracy of the simulated signals and the possible range for simulation were verified. It was found that, there is a good agreement between the CIVA simulated and experimental results in the A-scan signal, B-scan image, and measurement of depth.

  14. Simulation of time of flight defraction signals for reactor vessel head penetrations

    International Nuclear Information System (INIS)

    Lim, Tae Hun; Kim, Young Sik; Lee, Jeong Seok

    2016-01-01

    The simulation of nondestructive testing has been used in the prediction of the signal characteristics of various defects and in the development of the procedures. CIVA, a simulation tool dedicated to nondestructive testing, has good accuracy and speed, and provides a three-dimensional graphical user interface for improved visualization and familiar data displays consistent with an NDE technique. Even though internal validations have been performed by the CIVA software development specialists, an independent validation study is necessary for the assessment of the accuracy of the software prior to practical use. In this study, time of flight diffraction signals of ultrasonic inspection of a calibration block for reactor vessel head penetrations were simulated using CIVA. The results were compared to the experimentally inspected signals. The accuracy of the simulated signals and the possible range for simulation were verified. It was found that, there is a good agreement between the CIVA simulated and experimental results in the A-scan signal, B-scan image, and measurement of depth

  15. MAAP4 hot leg and lower head failure benchmarking

    International Nuclear Information System (INIS)

    Lee, S.J.; Henry, R.E.; Paik, C.Y.; Conzen, J.; Luangdilok, W.

    2009-01-01

    The MAAP4 material creep calculation was compared with the experiments reported by Maile, et al., for a 0.7 m diameter hot leg, with a thickness of 47 mm, which is pressurized to 16.3 MPa and heated to temperatures in excess of 700degC. These experiments showed that the carbon steel hot leg would undergo material creep to a failure state in approximately 1,100 seconds. In addition, the MAAP4 creep calculation was compared with the lower head failure tests performed at the Sandia National Laboratories (SNL). These experiments were performed using scaled models of a typical Reactor Pressure Vessel lower head. The test vessel was fabricated from SA533B1 steel with an inner diameter of 0.91 m and a nominal thickness of 30 mm. The experiments were performed at around 10 MPa internal pressure with various imposed heat flux distributions. The onset of creep was observed to occur between 660degC and 705degC. The MAAP4 model provides a good characterization of the material creep behavior. For the hot leg test benchmark, the key is determining the correct equivalent stress when the stress is multi-axial. A good agreement was obtained when a multiplier of 1.09 to the hoop stress was used. For the lower head failure benchmark, using correct creep properties is important. The SNL test vessel material was fabricated as SA533B1 steel. However, when the experimental vessel material was tested for creep properties it turned out to be significantly weaker than the reactor vessel steel which has the same identification. Also, the material undergoing phase transition and becoming stronger at high temperatures has to be considered for accurate prediction of the failure time. A good agreement was obtained when the creep data of Jeong, et al., was used. (author)

  16. 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

  17. 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

  18. Heat transfer between relocated materials and the RPV lower head

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Knudson, D.L. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States); Kohriyama, T. [INSS, Fukui (Japan)

    2001-07-01

    Questions about the coolability of a continuous mass of relocated corium were raised during the Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) Post-accident examinations indicate that nearly half of the material that relocated to the vessel lower head during the TMI-2 accident formed a cohesive or ''continuous'' layer. TMI-2 VIP results and other evidence suggest that conduction through this continuous layer of solidified corium materials was assisted by other cooling mechanisms. Because increased knowledge about in-vessel coolability of corium materials may assist reactor designers in demonstrating that their concepts are passively safe, there is international interest in this topic. However, data are needed to identify what cooling mechanism(s) occurred and to develop a validated model for predicting this cooling. Corium cooling models significantly impact predictions for subsequent accident progression, such as the estimated time and mode of vessel failure. Hence, improved cooling models will provide a much needed, missing component of severe accident analyses. This paper provides a critical review of research investigating the coolability of corium relocating to a water-filled lower head. Where possible, existing models and data for predicting cooling are quantitatively compared; and governing relationships are identified. Key phenomena that should be incorporated into models for predicting this heat transfer are discussed, and deficiencies in current models and available data for predicting cooling are noted. Recommendations for improving these models and for obtaining data to validate these models are also provided. (author)

  19. Computational simulation of natural convection of a molten core in lower head of a PWR pressure vessel

    International Nuclear Information System (INIS)

    Vieira, Camila Braga; Romero, Gabriel Alves; Jian Su

    2010-01-01

    Computational simulation of natural convection in a molten core during a hypothetical severe accident in the lower head of a typical PWR pressure vessel was performed for two-dimensional semi-circular geometry with isothermal walls. Transient turbulent natural convection heat transfer of a fluid with uniformly distributed volumetric heat generation rate was simulated by using a commercial computational fluid dynamics software ANSYS CFX 12.0. The Boussinesq model was used for the buoyancy effect generated by the internal heat source in the flow field. The two-equation k-ω based SST (Shear Stress Transport) turbulence model was used to mould the turbulent stresses in the Reynolds-Average Navier-Stokes equations (RANS). Two Prandtl numbers, 6:13 and 7:0, were considered. Five Rayleigh numbers were simulated for each Prandtl number used (109, 1010, 1011, 1012, and 1013). The average Nusselt numbers on the bottom surface of the semicircular cavity were in excellent agreement with Mayinger et al. (1976) correlation and only at Ra = 109 the average Nusselt number on the top flat surface was in agreement with Mayinger et al. (1976) and Kulacki and Emara (1975) correlations. (author)

  20. Hierarchical and coupling model of factors influencing vessel traffic flow.

    Science.gov (United States)

    Liu, Zhao; Liu, Jingxian; Li, Huanhuan; Li, Zongzhi; Tan, Zhirong; Liu, Ryan Wen; Liu, Yi

    2017-01-01

    Understanding the characteristics of vessel traffic flow is crucial in maintaining navigation safety, efficiency, and overall waterway transportation management. Factors influencing vessel traffic flow possess diverse features such as hierarchy, uncertainty, nonlinearity, complexity, and interdependency. To reveal the impact mechanism of the factors influencing vessel traffic flow, a hierarchical model and a coupling model are proposed in this study based on the interpretative structural modeling method. The hierarchical model explains the hierarchies and relationships of the factors using a graph. The coupling model provides a quantitative method that explores interaction effects of factors using a coupling coefficient. The coupling coefficient is obtained by determining the quantitative indicators of the factors and their weights. Thereafter, the data obtained from Port of Tianjin is used to verify the proposed coupling model. The results show that the hierarchical model of the factors influencing vessel traffic flow can explain the level, structure, and interaction effect of the factors; the coupling model is efficient in analyzing factors influencing traffic volumes. The proposed method can be used for analyzing increases in vessel traffic flow in waterway transportation system.

  1. SCDAP/RELAP5 Modeling of Movement of Melted Material Through Porous Debris in Lower Head

    International Nuclear Information System (INIS)

    Siefken, L. J.

    1998-01-01

    Designs are described for implementing models for calculating the movement of melted material through the interstices in a matrix of porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head during a severe accident in a Light Water Reactor. Currently, the COUPLE model has no capability to model the movement of material that melts within a matrix of porous material. The COUPLE model also does not have the capability to model the movement of liquefied core plate material that slumps onto a porous debris bed in the lower head. In order to advance beyond the assumption the liquefied material always remains stationary, designs are developed for calculations of the movement of liquefied material through the interstices in a matrix of porous material. Correlations are identified for calculating the permeability of the porous debris and for calculating the rate of flow of liquefied material through the interstices in the debris bed. Correlations are also identified for calculating the relocation of solid debris that has a large amount of cavities due to the flowing away of melted material. Equations are defined for calculating the effect on the temperature distribution in the debris bed of heat transported by moving material and for changes in effective thermal conductivity and heat capacity due to the movement of material. The implementation of these models is expected to improve the calculation of the material distribution and temperature distribution of debris in the lower head for cases in which the debris is porous and liquefied material is present within the porous debris

  2. Stress corrosion cracking in the vessel closure head penetrations of French PWR`s; Fissuration par corrosion sous contrainte de penetrations de couvercle de cuve de reacteur nucleaire francais a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Buisine, D.; Cattant, F.; Champredonde, J.; Pichon, C.; Benhamou, C.; Gelpi, A.; Vaindirlis, M.

    1994-01-01

    During a hydrotest in September 1991, part of the statutory decennial in-service inspection, a leak was detected on the vessel head of Bugey 3, which is one of the first 900 MW 3-loop PWR`s in France. This leak was due to a cracked penetration used for a control rod drive mechanism. The investigations performed identified Primary Stress Corrosion Cracking of Alloy 600 as being the origin of this degradation. So a lot of the same design PWR`s are a concern due to this generic problem. In this case, PWSCC was linked to: - hot temperature of the vessel head; - high residual stresses due to the welding process between peripherical penetrations and the vessel head; - sensitivity of forged Alloy 600 used for penetration manufacturing. This following paper will present the cracked analysis based, in particular, on the main results obtained in France on each of these items. These results come from the operating experience, the destructive examinations and the programs which are running on stress analysis and metallurgical characterizations. (authors). 9 figs., 2 tabs.

  3. Models and Algorithms for Container Vessel Stowage Optimization

    DEFF Research Database (Denmark)

    Delgado-Ortegon, Alberto

    .g., selection of vessels to buy that satisfy specific demands), through to operational decisions (e.g., selection of containers that optimize revenue, and stowing those containers into a vessel). This thesis addresses the question of whether it is possible to formulate stowage optimization models...... container of those to be loaded in a port should be placed in a vessel, i.e., to generate stowage plans. This thesis explores two different approaches to solve this problem, both follow a 2-phase decomposition that assigns containers to vessel sections in the first phase, i.e., master planning...

  4. Hierarchical and coupling model of factors influencing vessel traffic flow.

    Directory of Open Access Journals (Sweden)

    Zhao Liu

    Full Text Available Understanding the characteristics of vessel traffic flow is crucial in maintaining navigation safety, efficiency, and overall waterway transportation management. Factors influencing vessel traffic flow possess diverse features such as hierarchy, uncertainty, nonlinearity, complexity, and interdependency. To reveal the impact mechanism of the factors influencing vessel traffic flow, a hierarchical model and a coupling model are proposed in this study based on the interpretative structural modeling method. The hierarchical model explains the hierarchies and relationships of the factors using a graph. The coupling model provides a quantitative method that explores interaction effects of factors using a coupling coefficient. The coupling coefficient is obtained by determining the quantitative indicators of the factors and their weights. Thereafter, the data obtained from Port of Tianjin is used to verify the proposed coupling model. The results show that the hierarchical model of the factors influencing vessel traffic flow can explain the level, structure, and interaction effect of the factors; the coupling model is efficient in analyzing factors influencing traffic volumes. The proposed method can be used for analyzing increases in vessel traffic flow in waterway transportation system.

  5. A phenomenological analysis of melt progression in the lower head of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, J.M., E-mail: jean-marie.seiler@cea.fr [CEA, DEN, DTN, F-38054 Grenoble (France); Tourniaire, B. [EDF/Septen, Lyon (France)

    2014-03-15

    Highlights: • We propose a phenomenological description of melt progression into the lower head. • We examine changes in heat loads on the vessel. • Heat loads are more severe than emphasized by the bounding situation assumption. • Both primary circuit and ex-vessel reflooding are necessary for in-vessel retention. • Vessel failure conditions are examined. - Abstract: The analysis of in-vessel corium cooling (IVC) and retention (IVR) involves the description of very complex and transient physical phenomena. To get round this difficulty, “bounding” situations are often emphasized for the demonstration of corium coolability, by vessel flooding and/or by reactor pit flooding. This approach however comes up against its own limitations. More realistic melt progression scenarios are required to provide plausible corium configurations and vessel failure conditions. Work to develop more realistic melt progression scenarios has been done at CEA, in collaboration with EDF. Development has concentrated on the French 1300 MWe PWR, considering both dry scenarios and the possibility of flooding of the RPC (reactor primary circuit) and/or the reactor pit. The models used for this approach have been derived from the analysis of the TMI2 accident and take benefit from the lessons derived from several programs related to pool thermal hydraulics (BALI, COPO, ACOPO, etc.), material interactions (RASPLAV, MASCA), critical heat flux (CHF) on the external surface of the vessel (KAIST, SULTAN, ULPU), etc. Important conclusions of this work are as follows: (a)After the start of corium melting and onset of melt formation in the core at low pressure (∼1 to 5 bars), it seems questionable that RPV (reactor pressure vessel) reflooding alone would be sufficient to achieve corium retention in the vessel; (b)If the vessel is not cooled externally, it may fail due to local heat-up before the whole core fuel inventory is relocated in the lower head; (c)Even if the vessel is

  6. VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation)], E-mail: bechta@sbor.spb.su; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V.; Almiashev, V.I. [Institute of Silicate Chemistry, Russian Academy of Sciences (ISCh RAS), St. Petersburg (Russian Federation); Lopukh, D.B. [SPb State Electrotechnical University (SPbGETU), St. Petersburg (Russian Federation); Bottomley, D. [EUROPAISCHE KOMMISSION, Joint Research Centre Institut fuer Transurane (ITU), Karlsruhe (Germany); Fischer, M. [AREVA NP GmbH, Erlangen (Germany); Piluso, P. [CEA/DEN/DSNI, Saclay (France); Miassoedov, A.; Tromm, W. [Forschungszentrum Karlsruhe, Karlsruhe (Germany); Altstadt, E. [Forschungszentrum Rossendorf (FZR), Dresden (Germany); Fichot, F. [IRSN/DPAM/SEMCA, St. Paul lez Durance (France); Kymalainen, O. [FORTUM Nuclear Services Ltd., Espoo (Finland)

    2009-06-15

    The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations.

  7. VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere

    International Nuclear Information System (INIS)

    Bechta, S.V.; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A.; Gusarov, V.V.; Almiashev, V.I.; Lopukh, D.B.; Bottomley, D.; Fischer, M.; Piluso, P.; Miassoedov, A.; Tromm, W.; Altstadt, E.; Fichot, F.; Kymalainen, O.

    2009-01-01

    The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations.

  8. Lower head failure analysis

    International Nuclear Information System (INIS)

    Rempe, J.L.; Thinnes, G.L.; Allison, C.M.; Cronenberg, A.W.

    1991-01-01

    The US Nuclear Regulatory Commission is sponsoring a lower vessel head research program to investigate plausible modes of reactor vessel failure in order to determine (a) which modes have the greatest likelihood of occurrence during a severe accident and (b) the range of core debris and accident conditions that lead to these failures. This paper presents the methodology and preliminary results of an investigation of reactor designs and thermodynamic conditions using analytic closed-form approximations to assess the important governing parameters in non-dimensional form. Preliminary results illustrate the importance of vessel and tube geometrical parameters, material properties, and external boundary conditions on predicting vessel failure. Thermal analyses indicate that steady-state temperature distributions will occur in the vessel within several hours, although the exact time is dependent upon vessel thickness. In-vessel tube failure is governed by the tube-to-debris mass ratio within the lower head, where most penetrations are predicted to fail if surrounded by molten debris. Melt penetration distance is dependent upon the effective flow diameter of the tube. Molten debris is predicted to penetrate through tubes with a larger effective flow diameter, such as a boiling water reactor (BWR) drain nozzle. Ex-vessel tube failure for depressurized reactor vessels is predicted to be more likely for a BWR drain nozzle penetration because of its larger effective diameter. At high pressures (between ∼0.1 MPa and ∼12 MPa) ex-vessel tube rupture becomes a dominant failure mechanism, although tube ejection dominates control rod guide tube failure at lower temperatures. However, tube ejection and tube rupture predictions are sensitive to the vessel and tube radial gap size and material coefficients of thermal expansion

  9. A model for ultrasound contrast agent in a phantom vessel

    KAUST Repository

    Qamar, Adnan

    2014-02-01

    A theoretical framework to model the dynamics of Ultrasound Contrast Agent (UCA) inside a phantom vessel is presented. The model is derived from the reduced Navier-Stokes equation and is coupled with the evolving flow field solution inside the vessel by a similarity transformation approach. The results are computed, and compared with experiments available in literature, for the initial UCA radius of Ro=1.5 μm and 2 μm for the vessel diameter of D=12 μm and 200 μm with the acoustic parameters as utilized in the experiments. When compared to other models, better agreement on smaller vessel diameter is obtained with the proposed coupled model. The model also predicts, quite accurately, bubble fragmentation in terms of acoustic and geometric parameters. © 2014 IEEE.

  10. Structural failure analysis of reactor vessels due to molten core debris

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.

    1993-01-01

    Maintaining structural integrity of the reactor vessel during a postulated core melt accident is an important safety consideration in the design of the vessel. This paper addresses the failure predictions of the vessel due to thermal and pressure loadings from the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on a wet or dry cavity and pressurization of the vessel based on operating pressure or atmospheric (pipe break). The analyses considered both short term (minutes) and long term (days) failure modes. Short term failure modes include creep at elevated temperatures and plastic instabilities of the structure. Long term failure modes are caused by creep rupture that lead to plastic instability of the structure. The analyses predict the reactor vessel will remain intact after the core melt has deposited on the lower vessel head

  11. FOREVER Experiments on Thermal and Mechanical Behavior of a Reactor Pressure Vessel during a Severe Accident

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Nourgaliev, R.R.; Dinh, T.N.; Karbojian, A.; Green, J.A.; Bui, V.A.

    1999-01-01

    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 facility employs 1/10.-scale carbon steel vessels of 0.4 m diameter, 15 mm thickness and 600 mm height. Up to 20 liters of binary-oxide melts with 100-300 K superheat are employed, as a simulant for the prototypic corium melt, and internal heating is provided by electrical heaters of up to 20 kW power in order to maintain the vessel wall temperatures at 1100-1200 K. Auxiliary systems are designed to provide an overpressure up to 4 MPa in the test vessel. Thus, severe accident scenarios with RCS depressurization are modeled. Creep behavior of the three-dimensional vessel, formation of the gap between the melt pool crust and the creeping vessel, and mechanisms of the gap cooling by water ingression will be the subjects of study and measurements in the FOREVER experimental program. Scaling rationale as well as pre-test analyses of the thermal and mechanical behavior of the FOREVER test vessels are presented. (authors)

  12. A contrast enhancement and scanning techniques for CT angiography of head and neck. One phase injection method for simultaneous imaging of vessels and tumor

    International Nuclear Information System (INIS)

    Morita, Yasuhiko; Indo, Hiroko; Noikura, Takenori

    1999-01-01

    We report on a method of CT-Angiography useful for examining lesion of the head and neck using three-dimensional images and measured CT value. This study focused on some of the important blood vessels in the head and neck. The aim of this method was to obtain high-contrast enhancement for both vessels and tumors at same time. A total amount of 100 ml nonionic contrast media (Omnipaque 240, 240 mg iodine per milliliter, Daiichi seiyaku, Tokyo, Japan) was injected intravenously with a flow of 1.5 ml/sec. Spiral scans, 24 rotations with 24 seconds, were started at a time when remaining amount of contrast media had become 30 to 20 ml. All CT scans were performed using double speed spiral scan technique with a slice thickness of 2 to 3 mm and table speeds from 3 to 5 mm/rotation. The patients populations consisted of 9 men and 6 women who ranged in age from 37 to 85 years. Sixteen CT-angiography were performed according to this method. Mean CT values of major blood vessels were measured in order to find out threshold at the level of submandibular gland in 13 examinations for 12 subjects. Important vessels like the common, internal, and the external artery, internal and external jugular vein were clearly visible in all subjects. Three dimensional images of these vessels could also be reconstructed for 15 of the subjects. Mean CT values were 211 Hounsfield units (HU) and 209 HU for the right and left internal carotid artery, respectively, and 204 HU and 206 HU for the right and left external carotid artery, respectively. Mean CT values for right and left internal jugular vein were 195 HU and 194 HU respectively. Measured CT values at each important blood vessels showed this method could yields acceptable enhancements. Good enhancement effect of tumor and blood vessels in the same scan seems to be mutually incompatible. One very important trade-off is the early enhancement effect at blood vessels versus the late enhancement effect at tumors. The other important trade

  13. 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

  14. SCDAP/RELAP5 Modeling of Heat Transfer and Flow Losses in Lower Head Porous Debris

    International Nuclear Information System (INIS)

    Coryell, E.W.; Siefken, L.J.; Paik, S.

    1998-01-01

    Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and non-porous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of non-porous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate manner. A design is also described for implementing a model of heat transfer by radiation from debris to the interstitial fluid. A design is described for implementation of models for flow losses and interphase drag in porous debris. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region

  15. Synthetic analyses of the LAVA experimental results on in-vessel corium retention through gap cooling

    International Nuclear Information System (INIS)

    Kang, Kyoung Ho; Cho, Young Ro; Koo, Kil Mo; Park, Rae Joon; Kim, Jong Hwan; Kim, Jong Tae; Ha, Kwang Sun; Kim, Sang Baik; Kim, Hee Dong

    2001-03-01

    LAVA(Lower-plenum Arrested Vessel Attack) has been performed to gather proof of gap formation between the debris and lower head vessel and to evaluate the effect of the gap formation on in-vessel cooling. Through the total of 12 tests, the analyses on the melt relocation process, gap formation and the thermal and mechanical behaviors of the vessel were performed. The thermal behaviors of the lower head vessel were affected by the formation of the fragmented particles and melt pool during the melt relocation process depending on mass and composition of melt and subcooling and depth of water. During the melt relocation process 10.0 to 20.0 % of the melt mass was fragmented and also 15.5 to 47.5 % of the thermal energy of the melt was transferred to water. The experimental results address the non-adherence of the debris to the lower head vessel and the consequent gap formation between the debris and the lower head vessel in case there was an internal pressure load across the vessel abreast with the thermal load induced by the thermite melt. The thermal behaviors of the lower head vessel during the cooldown period were mainly affected by the heat removal characteristics through this gap, which were determined by the possibilities of the water ingression into the gap depending on the melt composition of the corium simulant. The enhanced cooling capacity through the gap was distinguished in the Al 2 O 3 melt tests. It could be inferred from the analyses on the heat removal capacity through the gap that the lower head vessel could effectively cooldown via heat removal in the gap governed by counter current flow limits(CCFL) even if 2mm thick gap should form in the 30 kg Al 2 O 3 melt tests, which was also confirmed through the variations of the conduction heat flux in the vessel and rapid cool down of the vessel outer surface in the Al 2 O 3 melt tests. In the case of large melt mass of 70 kg Al 2 O 3 melt, however, the infinite possibility of heat removal through the

  16. SCDAP/RELAP5 Modeling of Heat Transfer and Flow Losses in Lower Head Porous Debris

    International Nuclear Information System (INIS)

    Siefken, Larry James; Coryell, Eric Wesley; Paik, Seungho; Kuo, Han Hsiung

    1999-01-01

    Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of nonporous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate manner. Designs are described for models to calculate the flow losses and interphase drag of fluid flowing through the interstices of the porous debris, and to apply these variables in the momentum equations in the RELAP5 part of the code. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region

  17. Crack of reactor vessel upper head penetration nozzles in Korean nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Doh, E.; Lee, T-S.; Kim, J-Y.; Lee, C-H. [KEPCO Plant Service and Engineering Co., Ltd., Busan (Korea, Republic of)

    2014-07-01

    Since the first CRDM nozzles of reactor vessel head at Kori unit 1 in Korea were inspected in 2003, no CRDM nozzle cracks had been revealed prior to the inspection at Hanbit unit 3 in October 2012, even though many foreign plants had been reporting PWSCC cracks. In October 2012, seven axial cracks from 6 CRDM nozzles at Hanbit unit 3, and in November 2013, six axial cracks from 6 CRDM nozzles at Hanbit unit 4 were detected by TOFD Ultrasonic testing from ID of nozzles. There were confirmed to be PWSCC by Dye penetrant testing and Replica on the surface of J-groove weld of CRDM nozzles. Both plants are OPR-1000 types. All flaws started from the surface of J-groove weld at interface with OD of nozzle, but did not grow up to the top of J-groove weld, and did not make any Leak path up to head outside. The Performance Demonstration Initiative (PDI) system of CRDM nozzle inspection for Westinghouse type plants has been applied in Korea since July 2011. However, its application for OPR-1000 is still under development in Korea. The experience of PDI inspection for Westinghouse type plant contributed greatly to the detection and evaluation of PWSCC of CRDM nozzles at OPR- 1000 of Hanbit unit 3 & 4. The experimentally based procedure of flaw detection and the enhanced detection technique of examiners made it possible to detect and to determine the PWSCC indications. Embedded Flaw Repair process was approved by government authority, and the repair of the 6 CRDM nozzles in each plant was conducted by a consortium of Westinghouse and KPS. (author)

  18. [Duodenum-preserving total pancreatic head resection and pancreatic head resection with segmental duodenostomy].

    Science.gov (United States)

    Takada, Tadahiro; Yasuda, Hideki; Nagashima, Ikuo; Amano, Hodaka; Yoshiada, Masahiro; Toyota, Naoyuki

    2003-06-01

    A duodenum-preserving pancreatic head resection (DPPHR) was first reported by Beger et al. in 1980. However, its application has been limited to chronic pancreatitis because of it is a subtotal pancreatic head resection. In 1990, we reported duodenum-preserving total pancreatic head resection (DPTPHR) in 26 cases. This opened the way for total pancreatic head resection, expanding the application of this approach to tumorigenic morbidities such as intraductal papillary mucinous tumor (IMPT), other benign tumors, and small pancreatic cancers. On the other hand, Nakao et al. reported pancreatic head resection with segmental duodenectomy (PHRSD) as an alternative pylorus-preserving pancreatoduodenectomy technique in 24 cases. Hirata et al. also reported this technique as a new pylorus-preserving pancreatoduodenostomy with increased vessel preservation. When performing DPTPHR, the surgeon should ensure adequate duodenal blood supply. Avoidance of duodenal ischemia is very important in this operation, and thus it is necessary to maintain blood flow in the posterior pancreatoduodenal artery and to preserve the mesoduodenal vessels. Postoperative pancreatic functional tests reveal that DPTPHR is superior to PPPD, including PHSRD, because the entire duodenum and duodenal integrity is very important for postoperative pancreatic function.

  19. Re-expression of pro-fibrotic, embryonic preserved mediators in irradiated arterial vessels of the head and neck region

    Energy Technology Data Exchange (ETDEWEB)

    Moebius, Patrick; Preidl, Raimund H.M.; Weber, Manuel; Neukam, Friedrich W.; Wehrhan, Falk [Friedrich-Alexander-Universitaet Erlangen-Nuernberg (FAU), Department of Oral and Maxillofacial Surgery, University Hospital of Erlangen, Erlangen (Germany); Amann, Kerstin [Friedrich-Alexander-Universitaet Erlangen-Nuernberg (FAU), Department of Nephropathology, Institute of Pathology, University Hospital of Erlangen, Erlangen (Germany)

    2017-11-15

    Surgical treatment of head and neck malignancies frequently includes microvascular free tissue transfer. Preoperative radiotherapy increases postoperative fibrosis-related complications up to transplant loss. Fibrogenesis is associated with re-expression of embryonic preserved tissue developmental mediators: osteopontin (OPN), regulated by sex-determining region Y-box 9 (Sox9), and homeobox A9 (HoxA9) play important roles in pathologic tissue remodeling and are upregulated in atherosclerotic vascular lesions; dickkopf-1 (DKK1) inhibits pro-fibrotic and atherogenic Wnt signaling. We evaluated the influence of irradiation on expression of these mediators in arteries of the head and neck region. DKK1, HoxA9, OPN, and Sox9 expression was examined immunohistochemically in 24 irradiated and 24 nonirradiated arteries of the lower head and neck region. The ratio of positive cells to total cell number (labeling index) in the investigated vessel walls was assessed semiquantitatively. DKK1 expression was significantly decreased, whereas HoxA9, OPN, and Sox9 expression were significantly increased in irradiated compared to nonirradiated arterial vessels. Preoperative radiotherapy induces re-expression of embryonic preserved mediators in arterial vessels and may thus contribute to enhanced activation of pro-fibrotic downstream signaling leading to media hypertrophy and intima degeneration comparable to fibrotic development steps in atherosclerosis. These histopathological changes may be promoted by HoxA9-, OPN-, and Sox9-related inflammation and vascular remodeling, supported by downregulation of anti-fibrotic DKK1. Future pharmaceutical strategies targeting these vessel alterations, e. g., bisphosphonates, might reduce postoperative complications in free tissue transfer. (orig.) [German] Die operative Behandlung von Tumoren im Kopf- und Halsbereich umfasst den Transfer mikrovaskulaerer Gewebetransplantate. Praeoperative Bestrahlung verursacht eine erhoehte Inzidenz

  20. In-service ultrasonic inspection of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Prepechal, J.; Sulc, J.

    1982-01-01

    Ultrasonic tests of pressure vessels for WWER 440 reactors, type 213 V, are carried out partly manually and partly by test equipment. The inner surface of the pressure vessel is tested using device REACTORTEST TRC which is fully mobile. The outer surface of the cylindrical parts and bottoms of the body is tested using handling equipment permanently in-built under the pressure vessel and dismountable testing heads. A set of these heads may be used for two reactor units. The testing equipment REACTORTEST TRC is equipped with a TRC 800 ultrasound device. The equipment for testing the outer surface of the vessel operates with the UDAR 16 ultrasound apparatus to which may be simultaneously connected 10 ultrasound probes and six probes for acoustic feedback. The whole system of ultrasonic tests makes possible a first-rate and reliable volume control of the whole pressure vessel and all points where cracks may originate and grow. (Z.M.)

  1. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  2. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    International Nuclear Information System (INIS)

    Monteleone, S.

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  3. Radiation heat transfer in a pressurized water reactor lower head filled with molten corium

    International Nuclear Information System (INIS)

    Šadek, Siniša; Grgić, Davor; Debrecin, Nenad

    2013-01-01

    Highlights: ► We develop radiation heat exchange model for a reactor pressure vessel lower head. ► Model is used during a late in-vessel phase of severe accidents. ► View factors are calculated automatically for a time-dependent enclosure. ► Model is included in the RELAP5/SCDAPSIM computer code. ► Inclusion of heat radiation causes faster heat-up rate of RPV lower head structures. - Abstract: Following a core melt, molten material may slump to the lower head of a reactor pressure vessel (RPV). In that case, some structures like lower parts of fuel elements and a core support plate would remain intact. Since the melt is at high temperature and there are no obstacles between the melt and the supporting plate, the plate is exposed to an intense radiation heating. The radiation heat exchange model of the lower head was developed and applied to a finite element code COUPLE which is a part of the detailed mechanistic code RELAP5/SCDAPSIM. The radiation enclosure consisted of three surfaces: the upper surface of the relocated material, the inner surface of the RPV wall above the relocated material and the lower surface of the core support plate. View factors were calculated for the enclosure geometry that is changing in time because of intermittent accumulation of molten material. The enclosure surfaces were covered by mesh of polygonal areas and view factors were calculated, for each pair of the element areas, by solving the definite integrals using the algorithms for adaptive integrations by means of Gaussian quadrature. Algebraic equations for radiosity and irradiation vectors were solved by LU decomposition and the radiation model was explicitly coupled with the heat conduction model. The results show that there is a possibility of the core support plate failure after being heated up due to radiation heat exchange with the melt.

  4. Marine Vessel Models in Changing Operational Conditions - A Tutorial

    DEFF Research Database (Denmark)

    Perez, Tristan; Sørensen, Asgeir; Blanke, Mogens

    2006-01-01

    conditions (VOC). However, since marine systems operate in changing VOCs, there is a need to adapt the models. To date, there is no theory available to describe a general model valid across different VOCs due to the complexity of the hydrodynamic involved. It is believed that system identification could......This tutorial paper provides an introduction, from a systems perspective, to the topic of ship motion dynamics of surface ships. It presents a classification of parametric models currently used for monitoring and control of marine vessels. These models are valid for certain vessel operational...

  5. Accuracy of geometrical modelling of heat transfer from tissue to blood vessels

    NARCIS (Netherlands)

    Leeuwen, van G.M.J.; Kotte, A.N.T.J.; Bree, de J.; Koijk, van der J.F.; Crezee, J.; Lagendijk, J.J.W.

    1997-01-01

    We have developed a thermal model in which blood vessels are described as geometrical objects, 3D curves with associated diameters. Here the behaviour of the model is examined for low resolutions compared with the vessel diameter and for strongly curved vessels. The tests include a single straight

  6. Modeling irradiation embrittlement in reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Odette, G.R.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

  7. Analysis of Reactor Vessel Lower Head Penetration Tube Failure

    International Nuclear Information System (INIS)

    Stempniewicz, Marek

    1999-01-01

    This paper presents results of two studies, performed to investigate the behavior of the reactor vessel penetration tubes in case of relocation of molten material into the tubes. The first study is on the CORVIS drain line experiment 03/1. Results of pre-test calculations are presented, and compared to the later obtained experimental data. The timing of the drain line melting and the velocity of the debris flowing inside the drain line were predicted correctly, but the penetration depth was clearly underestimated. If the calculations are done using different correlation for the melt-to-wall convective heat transfer, the results are closer to the experiment. It cannot however be concluded that the alternative correlation is more appropriate until other uncertainties are clarified. The second study presents calculations performed for GKN Dodewaard CRD, instrument tubes and drain line. Calculations were performed to estimate whether the tubes have a chance to withstand the first attack of the melt and thus postpone vessel failure until the water in the lower plenum evaporates. Calculations were performed assuming that the melt can move into the tubes without any resistance, e.g. presence of water in the tubes was not taken into account. The results indicate that the critical penetration of the GKN vessel, which is most likely to fail, is the drain line. Results also indicate that external flooding should prevent early tube failure, at least in case of low vessel pressure. (author)

  8. Numerical investigation of the reactor pressure vessel behaviour under severe accident conditions taking into account the combined processes of the vessel creep and the molten pool natural convection

    International Nuclear Information System (INIS)

    Loktionov, V.D.; Mukhtarov, E.S.; Yaroshenko, N.I.; Orlov, V.E.

    1999-01-01

    Analysis of the WWER lower head behaviour and its failure has been performed for several molten pool structures and internal overpressure levels in a reactor pressure vessel (RPV). The different types of the molten pools (homogeneous, conventionally homogeneous, conventionally stratified, stratified) cover the bounding scenarios during a hypothetical severe accident. The parametric investigations of the failure mode and RPV behaviour for various molten pool types, its heights and internal overpressure levels are presented herein. A coupled treatment in this investigation includes: (i) a 2-D thermohydraulic analysis of a molten pool natural convection. Domestic NARAUFEM code has been used in this detailed analysis for prediction of the heat flux from the molten pool to the RPV inner surface; and (ii) a detailed 3-D transient thermal analysis of the RPV lower head. Domestic 3-D ASHTER-VVR finite element code has been used for the numerical simulations of the high temperature creep and failure of the lower head. The effect of an external RPV cooling, temperature-dependent physical properties of the molten pool and vessel steel, the hydrostatic forces and vessel dead-weight were taken into account in this study. The obtained results show that lower head failure occurs as a result of the vessel creep process which is significantly dependent on both an internal overpressure level and the type of molten pool structure. In particular, it was found that there were combinations of 'overpressure-molten pool structure' when the vessel failure started at the 'hot' layers of the vessel. (orig.)

  9. Predictive simulation of bidirectional Glenn shunt using a hybrid blood vessel model.

    Science.gov (United States)

    Li, Hao; Leow, Wee Kheng; Chiu, Ing-Sh

    2009-01-01

    This paper proposes a method for performing predictive simulation of cardiac surgery. It applies a hybrid approach to model the deformation of blood vessels. The hybrid blood vessel model consists of a reference Cosserat rod and a surface mesh. The reference Cosserat rod models the blood vessel's global bending, stretching, twisting and shearing in a physically correct manner, and the surface mesh models the surface details of the blood vessel. In this way, the deformation of blood vessels can be computed efficiently and accurately. Our predictive simulation system can produce complex surgical results given a small amount of user inputs. It allows the surgeon to easily explore various surgical options and evaluate them. Tests of the system using bidirectional Glenn shunt (BDG) as an application example show that the results produc by the system are similar to real surgical results.

  10. Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core Concrete Interaction

    International Nuclear Information System (INIS)

    Robb, Kevin R; Farmer, Mitchell; Francis, Matthew W

    2015-01-01

    Lower head failure and corium concrete interaction were predicted to occur at Fukushima Daiichi Unit 1 (1F1) by several different system-level code analyses, including MELCOR v2.1 and MAAP5. Although these codes capture a wide range of accident phenomena, they do not contain detailed models for ex-vessel core melt behavior. However, specialized codes exist for analysis of ex-vessel melt spreading (e.g., MELTSPREAD) and long-term debris coolability (e.g., CORQUENCH). On this basis, an analysis was carried out to further evaluate ex-vessel behavior for 1F1 using MELTSPREAD and CORQUENCH. Best-estimate melt pour conditions predicted by MELCOR v2.1 and MAAP5 were used as input. MELTSPREAD was then used to predict the spatially dependent melt conditions and extent of spreading during relocation from the vessel. The results of the MELTSPREAD analysis are reported in a companion paper. This information was used as input for the long-term debris coolability analysis with CORQUENCH.

  11. Modelling of hydrogen deflagration in vessels using GOTHIC

    International Nuclear Information System (INIS)

    Wang, L.L.; Wong, R.C.; Fluke, R.J.

    1997-01-01

    Simulations of hydrogen deflagration tests were performed using the discrete lumpedparameter bum model of the computer code GOTHIC. The tests were performed in small and large scale spherical vessels and a cylindrical vessel. The small vessel cases included the effects of venting, and the cylindrical tests included the effects of obstacles. The simulations were performed by sub-dividing the volumes into either five or ten 'cells', and parameters such as flame speed and hydrogen concentration were varied. Measured flame speeds were used in the simulations and the results were compared to simulations using the code 'default' flame speed. The calculated pressure transients compared well with the experimental results using the measured flame speeds in the simulations of unvented cases, whereas for vented cases, the predicted peak pressures were generally less than the measurements. However, when the code default flame speed is used, the predicted peak pressures were more consistent and generally conservative when compared with the measurements. When the default flame speeds were used for vessels without obstacles, the peak pressures obtained were higher and the bum times were shorter than the experimental measurements. This was probably due to the basis for the correlations used for default flame speed in the bum model. These correlations were derived from intermediate-scale experiments for hydrogen combustion in relatively turbulent (fans on) environments. For vessels without obstacles, laminar flame speeds were more likely. Hence, the predicted peak pressures would be expected to be higher than the experimental results. In order to account for the degree of turbulence and flame acceleration caused by the presence of obstacles, higher than default flame speeds were used in the simulation of the vessel with obstacles. It was found that twice the default flame speed provided predictions of peak pressures comparable to the measurements. Based on the simulations conducted

  12. Design Procedure on Stud Bolt for Reactor Vessel Assembly

    International Nuclear Information System (INIS)

    Kim, Jong-Wook; Lee, Gyu-Mahn; Jeoung, Kyeong-Hoon; Kim, Tae-Wan; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-01

    The reactor pressure vessel flange is welded to the upper part of reactor pressure vessel, and there are stud holes to mount the closure head with stud bolts. The surface mating the closure head is compressed with O-ring, which acts as a sealing gasket to prevent coolant leakage. Bolted flange connections perform a very important structural role in the design of a reactor pressure vessel. Their importance stems from two important functions: (a) maintenance of the structural integrity of the connection itself, and (b) prevention of leakage through the O-ring preloaded by stud bolts. In the present study, an evaluation procedure for the design of stud bolt is developed to meet ASME code requirements. The developed design procedure could provide typical references in the development of advanced reactor design in the future

  13. SCDAP/RELAP5 modeling of heat transfer and flow losses in lower head porous debris. Rev. 1

    International Nuclear Information System (INIS)

    Siefken, L.J.; Coryell, E.W.; Paik, S.; Kuo, H.

    1999-01-01

    Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of nonporous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate ma nner. Designs are described for models to calculate the flow losses and interphase drag of fluid flowing through the interstices of the porous debris, and to apply these variables in the momentum equations in the RELAP5 part of the code. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region

  14. Fast-neutron nuclear reactor vessel

    International Nuclear Information System (INIS)

    Presciuttini, L.

    1984-01-01

    The reactor vessel comprises a cylindrical shell, of which axis is vertical, coupled at its lower part with a dished bottom. The reactor core rests on a support plate bearing on the bottom of the vessel. The above dished bottom comprises a spherical sector having the same radius and the same axis as the cylindrical shell and joining the lower part of the shell, and a spherical head of which radius is a little more important than the spherical sector one. A cylindrical support for the reactor core is attached to the vessel at the joint between the two dished sections. The invention applies more particularly to integrated type reactors cooled by liquid sodium [fr

  15. 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.

  16. GAREC analyses in Support of In-Vessel Retention Concept

    International Nuclear Information System (INIS)

    Azarian, G.; Gandrille, P.; Dumontet, A.; Grange; Barbier, F; Bellon, M.; Bordier, G.; Boulanger, F.; Cognet, G.; Gatt, J.M.; Humbert, J.M.; Laporte, T.; Lepareux, M.; Richard, P.; Robert, G.; Seiler, J.M.; Szabo, I.; Tourasse, M.; Valin, F.; Van Dorsselaere, J.P.

    1999-01-01

    The authors describe the analyses of the in-vessel retention capability which the GAREC group has performed for present and future French PWR designs. They present the reactor characteristics which are considered, describe the physical situations which are analysed and the relocation processes initiated by a corium flow, discuss the jet impacts, the debris formation and behaviour in the vessel lower head in a dry situation with absence of cooling, in wet situations in absence of external cooling, in wet situation with external cooling, in dry situation with external cooling. In this last case, they discuss the power dissipated in the corium, the molten salt behaviour, the heat flux distribution from the pool, the residual wall thickness, the heat flux distribution from the metal layer, the thermal-hydraulic aspects of water injection in the pool, the effects of crust instabilities, the external cooling, and the vessel mechanical behaviour. Then, they address the vapour explosion which may occur: mechanical loads leading to vessel failure in the cases of an eroded or non-eroded vessel, corium masses participating to the interaction (corium jets to the lower head, reflooding of corium pools with water). They finally briefly discuss the possible design improvements for in-vessel retention

  17. Prestressed pressure vessel for nuclear power plants

    International Nuclear Information System (INIS)

    1974-01-01

    The pressure vessel consists of a wall, a bottom, and a closure head, the wall being composed of annular segments. The closure head can be seated on the edge of the wall. Wall and closure head have got axial prestressing channels in which through-going steel tendons are arranged. They are concentrated in bundles and held above the head by anchoring devices. Within the prestressing channels of the head there are supporting jackets attached to the edge of the wall and projecting from the head through a coller. The anchoring devices, e.g. anchoring plates, may be optionally supported on the collars of the supporting jackets or on the closure head by means of auxiliary devices. The auxiliary devices for this purpose consist of extension nuts attached to the anchoring plates and closure head connecting shells. The closure head therefore may be drawn off over the anchoring devices. (DG) [de

  18. Modeling and measurement of the motion of the DIII-D vacuum vessel during vertical instabilities

    International Nuclear Information System (INIS)

    Reis, E.; Blevins, R.D.; Jensen, T.H.; Luxon, J.L.; Petersen, P.I.; Strait, E.J.

    1991-11-01

    The motions of the D3-D vacuum vessel during vertical instabilities of elongated plasmas have been measured and studied over the past five years. The currents flowing in the vessel wall and the plasma scrapeoff layer were also measured and correlated to a physics model. These results provide a time history load distribution on the vessel which were input to a dynamic analysis for correlation to the measured motions. The structural model of the vessel using the loads developed from the measured vessel currents showed that the calculated displacement history correlated well with the measured values. The dynamic analysis provides a good estimate of the stresses and the maximum allowable deflection of the vessel. In addition, the vessel motions produce acoustic emissions at 21 Hertz that are sufficiently loud to be felt as well as heard by the D3-D operators. Time history measurements of the sounds were correlated to the vessel displacements. An analytical model of an oscillating sphere provided a reasonable correlation to the amplitude of the measured sounds. The correlation of the theoretical and measured vessel currents, the dynamic measurements and analysis, and the acoustic measurements and analysis show that: (1) The physics model can predict vessel forces for selected values of plasma resistivity. The model also predicts poloidal and toroidal wall currents which agree with measured values; (2) The force-time history from the above model, used in conjunction with an axisymmetric structural model of the vessel, predicts vessel motions which agree well with measured values; (3) The above results, input to a simple acoustic model predicts the magnitude of sounds emitted from the vessel during disruptions which agree with acoustic measurements; (4) Correlation of measured vessel motions with structural analysis shows that a maximum vertical motion of the vessel up to 0.24 in will not overstress the vessel or its supports. 11 refs., 10 figs., 1 tab

  19. Kinematics of a Head-Neck Model Simulating Whiplash

    Science.gov (United States)

    Colicchia, Giuseppe; Zollman, Dean; Wiesner, Hartmut; Sen, Ahmet Ilhan

    2008-01-01

    A whiplash event is a relative motion between the head and torso that occurs in rear-end automobile collisions. In particular, the large inertia of the head results in a horizontal translation relative to the thorax. This paper describes a simulation of the motion of the head and neck during a rear-end (whiplash) collision. A head-neck model that…

  20. OECD/CSNI Workshop on In-Vessel Core Debris Retention and Coolability - Summary and Conclusions

    International Nuclear Information System (INIS)

    Behbahani, Ali-Reza; Drozd, Andrzej; Kim, Sang-Baik; Micaelli, Jean-Claude; Okkonen, Timo; Sugimoto, Jun; Trambauer, Klaus; Tuomisto, Harri

    1999-01-01

    In the spring of 1994 an OECD Workshop on Large Pool Heat transfer was held in Grenoble. The scope of this workshop was the investigation of (1) molten pool heat transfer, (2) heat transfer to the surrounding water, and (3) the feasibility of in-vessel core debris cooling through external cooling of the vessel. Since this time, experimental test series have been completed (e.g., COPO, ULPU, CORVIS) and new experimental programs (e.g., BALI, SONATA, RASPLAV, debris and gap heat transfer) have been established to consolidate and expand the data base for further model development and to improve the understanding of in-vessel debris retention and coolability in a nuclear power plant. Discussions within the CSNI's PWG-2 and the Task Group on Degraded Core Cooling (TG-DCC) have led to the conclusion that the time was ripe for organizing a new international Workshop with the objectives: - to review the results of experimental research that has been conducted in this area; - to exchange information on the results of member countries experiments and model development on in-vessel core debris retention and coolability; - to discuss areas where additional experimental research is needed in order to provide an adequate data base for analytical model development for core debris retention and coolability. The scope of this workshop was limited to the phenomena connected to in-vessel core debris retention and coolability and did not include steam explosion and fission product issues. The workshop was structured into the following sessions: Key note papers; Experiments and model development; Debris bed heat transfer; Corium properties, molten pool convection and crust formation; Gap formation and gap cooling; Creep behaviour of reactor pressure vessel lower head; Ex-vessel boiling and critical heat flux phenomena; Scaling to reactor severe accident conditions and reactor applications. Compared to the previous workshop held in Grenoble in 1994, large progress has been made in the

  1. Stress analyses for reactor pressure vessels by the example of a product line '69 boiling water reactor

    International Nuclear Information System (INIS)

    Mkrtchyan, Lilit; Schau, Henry; Wolf, Werner; Holzer, Wieland; Wernicke, Robert; Trieglaff, Ralf

    2011-01-01

    The reactor pressure vessels (RPV) of boiling water reactors (BWR) belonging to the product line '69 have unusually designed heads. The spherical cap-shaped bottom head of the vessel is welded directly to the support flange of the lower shell course. This unusual construction has led repeatedly to controversial discussions concerning the limits and admissibility of stress intensities arising in the junction of the bottom head to the cylindrical shell. In the present paper, stress analyses for the design conditions are performed with the finite element method in order to determine and categorize the occurring stresses. The procedure of stress classification in accordance with the guidelines of German KTA 3201.2 and Section III of the ASME Code (Subsection NB) is described and subsequently demonstrated by the example of a typical BWR vessel. The accomplished investigations yield allowable stress intensities in the considered area. Additionally, limit load analyses are carried out to verify the obtained results. Complementary studies, performed for a torispherical head, prove that the determined maximum peak stresses in the junction between the bottom head and the cylindrical shell are not unusual also for pressure vessels with regular bottom head constructions. (orig.)

  2. [Effect of vascular endothelial growth factor and tumor necrosis factor receptor for treatment of avascular necrosis of the femoral head in rabbits].

    Science.gov (United States)

    Hu, Zhi-ming; Zhou, Ming-qian; Gao, Ji-min

    2008-12-01

    To evaluate the therapeutic effect of vascular endothelial growth factor (VEGF) and tumor necrosis factor receptor (TNFR) on avascular necrosis of the femoral head in rabbits. Avascular necrosis of the femoral head was induced in 26 New Zealand white rabbits by injections of horse serum and prednisolone. The rabbits were then divided into VEGF/TNFR treatment group, VEGF treatment group, and untreated model group, with another 4 normal rabbits as the normal control group. In the two treatment groups, the therapeutic agents were injected percutaneously into the femoral head. Enzyme-linked immunosorbent assay was performed to determine the concentration of TNF-alpha in rabbit serum followed by pathological examination of the changes in the bone tissues, bone marrow hematopoietic tissue and the blood vessels in the femoral head. Compared with the model group, the rabbits with both VEGF and TNFR treatment showed decreased serum concentration of TNF-alpha with obvious new vessel formation, decreased empty bone lacunae in the femoral head and hematopoietic tissue proliferation in the bone marrow cavity. Percutaneous injection of VEGF and TNFR into the femoral head can significantly enhance bone tissue angiogenesis and ameliorate osteonecrosis in rabbits with experimental femoral head necrosis.

  3. TMI-2 Vessel Investigation Project integration report

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, J. R.; Rempe, J. L.; Stickler, L. A.; Korth, G. E.; Diercks, D. R.; Neimark, L. A.; Akers, D W; Schuetz, B. K.; Shearer, T L; Chavez, S. A.; Thinnes, G. L.; Witt, R. J.; Corradini, M L; Kos, J. A. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-03-01

    The Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) was an international effort that was sponsored by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. The primary objectives of the VIP were to extract and examine samples from the lower head and to evaluate the potential modes of failure and the margin of structural integrity that remained in the TMI-2 reactor vessel during the accident. This report presents a summary of the major findings and conclusions that were developed from research during the VIP. Results from the various elements of the project are integrated to form a cohesive understanding of the vessel`s condition after the accident.

  4. The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

    2001-03-01

    In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary.

  5. The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

    International Nuclear Information System (INIS)

    Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

    2001-03-01

    In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary

  6. Polymer-based blood vessel models with micro-temperature sensors in EVE

    Science.gov (United States)

    Mizoshiri, Mizue; Ito, Yasuaki; Hayakawa, Takeshi; Maruyama, Hisataka; Sakurai, Junpei; Ikeda, Seiichi; Arai, Fumihito; Hata, Seiichi

    2017-04-01

    Cu-based micro-temperature sensors were directly fabricated on poly(dimethylsiloxane) (PDMS) blood vessel models in EVE using a combined process of spray coating and femtosecond laser reduction of CuO nanoparticles. CuO nanoparticle solution coated on a PDMS blood vessel model are thermally reduced and sintered by focused femtosecond laser pulses in atmosphere to write the sensors. After removing the non-irradiated CuO nanoparticles, Cu-based microtemperature sensors are formed. The sensors are thermistor-type ones whose temperature dependences of the resistance are used for measuring temperature inside the blood vessel model. This fabrication technique is useful for direct-writing of Cu-based microsensors and actuators on arbitrary nonplanar substrates.

  7. Stress analyses for reactor pressure vessels by the example of a product line '69 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mkrtchyan, Lilit; Schau, Henry [TUEV SUED Energietechnik GmbH, Mannheim (Germany). Abt. Strukturverhalten; Wolf, Werner; Holzer, Wieland [TUEV SUED Industrie Service GmbH, Muenchen (Germany). Abt. Behaelter und Turbosatz; Wernicke, Robert; Trieglaff, Ralf [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany). Abt. Festigkeit und Konstruktion

    2011-08-15

    The reactor pressure vessels (RPV) of boiling water reactors (BWR) belonging to the product line '69 have unusually designed heads. The spherical cap-shaped bottom head of the vessel is welded directly to the support flange of the lower shell course. This unusual construction has led repeatedly to controversial discussions concerning the limits and admissibility of stress intensities arising in the junction of the bottom head to the cylindrical shell. In the present paper, stress analyses for the design conditions are performed with the finite element method in order to determine and categorize the occurring stresses. The procedure of stress classification in accordance with the guidelines of German KTA 3201.2 and Section III of the ASME Code (Subsection NB) is described and subsequently demonstrated by the example of a typical BWR vessel. The accomplished investigations yield allowable stress intensities in the considered area. Additionally, limit load analyses are carried out to verify the obtained results. Complementary studies, performed for a torispherical head, prove that the determined maximum peak stresses in the junction between the bottom head and the cylindrical shell are not unusual also for pressure vessels with regular bottom head constructions. (orig.)

  8. Programmable - logic equipment for ultrasound periodic inspections of reactor pressure vessels

    International Nuclear Information System (INIS)

    Haniger, L.

    1980-01-01

    Two alternatives are presented of programmable logic corresponding to the 2nd generation of the apparatus for performing periodic ultrasonic inspections of power reactor pressure vessels and a solution is outlined of inspecting the circumferential weld on the pressure vessel head. The apparatus will allow using any measuring head taken into consideration for operational inspection. Command words are taken from a punched type reader. Czechoslovak made RAM memories are used. The algorithm of instrument function is supposed to be controlled by a microprocessor as soon as necessary preconditions for this technology are created in Czechoslovakia

  9. Design of pressure vessels using shape optimization: An integrated approach

    Energy Technology Data Exchange (ETDEWEB)

    Carbonari, R.C., E-mail: ronny@usp.br [Department of Mechatronic Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil); Munoz-Rojas, P.A., E-mail: pablo@joinville.udesc.br [Department of Mechanical Engineering, Universidade do Estado de Santa Catarina, Bom Retiro, Joinville, SC 89223-100 (Brazil); Andrade, E.Q., E-mail: edmundoq@petrobras.com.br [CENPES, PDP/Metodos Cientificos, Petrobras (Brazil); Paulino, G.H., E-mail: paulino@uiuc.edu [Newmark Laboratory, Department of Civil and Environmental Engineering, University of Illinois at Urbana-Champaign, 205 North Mathews Av., Urbana, IL 61801 (United States); Department of Mechanical Science and Engineering, University of Illinois at Urbana-Champaign, 158 Mechanical Engineering Building, 1206 West Green Street, Urbana, IL 61801-2906 (United States); Nishimoto, K., E-mail: knishimo@usp.br [Department of Naval Architecture and Ocean Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil); Silva, E.C.N., E-mail: ecnsilva@usp.br [Department of Mechatronic Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil)

    2011-05-15

    Previous papers related to the optimization of pressure vessels have considered the optimization of the nozzle independently from the dished end. This approach generates problems such as thickness variation from nozzle to dished end (coupling cylindrical region) and, as a consequence, it reduces the optimality of the final result which may also be influenced by the boundary conditions. Thus, this work discusses shape optimization of axisymmetric pressure vessels considering an integrated approach in which the entire pressure vessel model is used in conjunction with a multi-objective function that aims to minimize the von-Mises mechanical stress from nozzle to head. Representative examples are examined and solutions obtained for the entire vessel considering temperature and pressure loading. It is noteworthy that different shapes from the usual ones are obtained. Even though such different shapes may not be profitable considering present manufacturing processes, they may be competitive for future manufacturing technologies, and contribute to a better understanding of the actual influence of shape in the behavior of pressure vessels. - Highlights: > Shape optimization of entire pressure vessel considering an integrated approach. > By increasing the number of spline knots, the convergence stability is improved. > The null angle condition gives lower stress values resulting in a better design. > The cylinder stresses are very sensitive to the cylinder length. > The shape optimization of the entire vessel must be considered for cylinder length.

  10. Study on external reactor vessel cooling capacity for advanced large size PWR

    International Nuclear Information System (INIS)

    Jin Di; Liu Xiaojing; Cheng Xu; Li Fei

    2014-01-01

    External reactor vessel cooling (ERVC) is widely adopted as a part of in- vessel retention (IVR) in severe accident management strategies. In this paper, some flow parameters and boundary conditions, eg., inlet and outlet area, water inlet temperature, heating power of the lower head, the annular gap size at the position of the lower head and flooding water level, were considered to qualitatively study the effect of them on natural circulation capacity of the external reactor vessel cooling for an advanced large size PWR by using RELAP5 code. And the calculation results provide some basis of analysis for the structure design and the following transient response behavior of the system. (authors)

  11. 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

  12. Thermal Load Analysis of Multilayered Corium in the Lower Head of Reactor Pressure Vessel during Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Whang, Seok Won; Park, Hyun Sun [POSTECH, Pohang (Korea, Republic of); Hwang, Tae Suk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-05-15

    In-Vessel Retention (IVR) is one of the severe accident management strategies to terminate or mitigate the severe accident which is also called 'core-melt accident'. The reactor vessel would be cooled by flooding the cavity with water. The molten core mixture is divided into two or three layers due to the density difference. Light metal layer which contains Fe and Zr is on the oxide layer which is consist of UO{sub 2} and ZrO{sub 2}. Heavy metal layer which contains U, Fe and Zr is located under the oxide layer. In oxide layer, the crust which is solidified material is formed along the boundary. The assessment of IVR for nuclear power plant has been conducted with lumped parameter method by Theofanous, Rempe and Esmaili. In this paper, the numerical analysis was performed and verified with the Esmaili's work to analyze thermal load of multilayered corium in pressurized reactor vessel and also to examine the condition of in-vessel corium characteristic before the vessel failure that lead to ex-vessel severe accident progression for example, ex-vessel debris bed cooling. The in-vessel coolability analysis for several scenarios is conducted for the plant which has higher power than AP1000. Two sensitivity analyses are conducted, the first is emissivity of light metal layer and the second is the heat transfer coefficient correlations of oxide layer. The effect of three layered system also investigated. In this paper, the numerical analysis was performed and verified with Esmaili's model to analyze thermal load of multilayered corium in pressurized reactor vessel. For two layered system, thermal load was analyzed according to the severe accident scenarios, emissivity of the light metal layer and heat transfer correlations of the.

  13. Safety assessment of in-vessel vapor explosion loads in next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Kwang Hyun; Cho, Jong Rae; Choi, Byung Uk; Kim, Ki Yong; Lee, Kyung Jung [Korea Maritime University, Busan (Korea); Park, Ik Kyu [Seoul National University, Seoul (Korea)

    1998-12-01

    A safety assessment of the reactor vessel lower head integrity under in-vessel vapor explosion loads has been performed. The premixing and explosion calculations were performed using TRACER-II code. Using the calculated explosion pressures imposed on the lower head inner wall, strain calculations were performed using ANSYS code. The explosion analyses show that the explosion impulses are not altered significantly by the uncertain parameters of triggering location and time, fuel and vapor volume fractions in uniform premixture bounding calculations within the conservative ranges. Strain analyses using the calculated pressure loads on the lower head inner wall show that the vapor explosion-induced lower head failure is physically unreasonable. The static analysis using the conservative explosion-end pressure of 7,246 psia shows that the maximum equivalent strain is 4.3% at the bottom of lower head, which is less than the allowable threshold value of 11%. (author). 24 refs., 40 figs., 3 tabs.

  14. 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)

  15. Lymphatic vessel density in vocal cord carcinomas assessed with LYVE-1 receptor expression

    International Nuclear Information System (INIS)

    Koskinen, Walter J.; Bono, Petri; Leivo, Ilmo; Vaheri, Antti; Aaltonen, Leena-Maija; Joensuu, Heikki

    2005-01-01

    Background and purpose: Early stage vocal cord carcinomas are usually cured by radiation therapy despite the use of portals that exclude the cervical lymph nodes. We investigated whether the lymphatic vessel density (LVD) of vocal cord carcinomas differs from that of other head and neck carcinomas. Patients and methods: Deparaffinized tissue from tumors of 60 patients diagnosed with head and neck squamous cell carcinoma (HNSCC) were immunostained for LYVE-1, a novel lymphatic vessel marker. Twenty-two had vocal cord carcinoma. Tumor blood vessel density (BVD) was assessed using immunostaining for CD31. Results: Tumor overall LVD, including both intra- and peritumoral lymph vessels, was 10-fold lower than the BVD (5 counts/mm 2 vs. 52 mm -2 , respectively). A high LVD was associated with a high BVD (P=.002), but neither was associated with the tumor size. Both tumor LVD and BVD were lower in vocal cord carcinomas than in HNSCCs arising at other sites (median, 0 vs. 7 mm -2 , P=.016; and median, 36 vs. 52 mm -2 , P=.006, respectively). Only one vocal cord carcinoma was associated with a regional metastasis at the time of the diagnosis. Among the rest of the cases tumor size was a better predictor for the presence of regional metastases than tumor BVD or LVD in a logistic regression model (odds ratio 2.2, 95% CI 1.1-4.5). Conclusion: Vocal cord carcinomas have a low lymph vessel density as compared with HNSCCs arising at other sites

  16. On the prediction of the reactor vessel integrity under severe accident loadings (RPVSA)

    Energy Technology Data Exchange (ETDEWEB)

    Krieg, R. E-mail: maeule@irs.fzk.de; Devos, J.; Caroli, C.; Solomos, G.; Ennis, P.J.; Kalkhof, D

    2001-11-01

    In order to allow more reliable predictions on the lower head response under core melt-down conditions, the temperature distribution has been analysed including the natural convection in the corium pool. Furthermore, the mechanical models and the failure criteria have been improved based on the RUPTHER and FASTHER experiments where typical temperature gradients are simulated. Lower head local melting as well as corium crust development has been addressed in the CORVIS experiments studying the contact between an alumina/iron thermite and a thick steel plate. The upper head loading by corium impact due to a postulated in-vessel steam explosion has been investigated by the BERDA experiments. Similarity rules were considered such that the results can be directly converted to reactor conditions. Based on these investigations admissible steam explosion energy releases are determined which the upper head can carry. If these limits are not exceeded the reactor containment cannot be endangered by broken head fragments. To provide the necessary basic data, mechanical material tests have been performed.

  17. Adenocarcinoma of the pancreatic head: preoperative helical CT. Criteria of resectability

    International Nuclear Information System (INIS)

    Kozima, Shigeru; Szelagowski, Carlos; Tisserand, Guy L.; Ocampo, Carlos; Zandalazini, Hugo; Silva, Walter; Oria, Alejandro; Vidovic, Gustavo; Varas, Pablo

    2001-01-01

    Objective: The purpose of this study is to determine the accuracy of biphasic helical CT scanning in predicting resectability of adenocarcinoma of the head of the pancreas by staying tumor involvement of the portal and superior mesenteric veins. Material and methods: 46 patients with proven adenocarcinoma of the head of the pancreas who underwent curative or palliative surgery were studied with preoperative biphasic helical CT scanning. Tumor involvement of the portal and mesenteric veins was graduated on a 1-3 scale based on circumferential contiguity of the tumor vessel. Grade 1: without contact; grade 2: tumor involvement of less than 50% of the vessel; grade 3: tumor involvement of more than 50%. Results: The total number of vessels evaluated was 92. In our series the preoperative biphasic helical CT was accurate in 77% for resectability and unresectability. Conclusion: Our experience of staging in 3 grades with biphasic helical CT, vessel involvement the portal and superior mesenteric veins of adenocarcinoma of the head of the pancreas is highly specific for unresectable tumor in patients who were graded 2 and 3. (author)

  18. Stress categorization in nozzle to pressure vessel connections finite elements models

    International Nuclear Information System (INIS)

    Albuquerque, Levi Barcelos de

    1999-01-01

    The ASME Boiler and Pressure Vessel Code, Section III , is the most important code for nuclear pressure vessels design. Its design criteria were developed to preclude the various pressure vessel failure modes throughout the so-called 'Design by Analysis', some of them by imposing stress limits. Thus, failure modes such as plastic collapse, excessive plastic deformation and incremental plastic deformation under cyclic loading (ratchetting) may be avoided by limiting the so-called primary and secondary stresses. At the time 'Design by Analysis' was developed (early 60's) the main tool for pressure vessel design was the shell discontinuity analysis, in which the results were given in membrane and bending stress distributions along shell sections. From that time, the Finite Element Method (FEM) has had a growing use in pressure vessels design. In this case, the stress results are neither normally separated in membrane and bending stress nor classified in primary and secondary stresses. This process of stress separation and classification in Finite Element (FE) results is what is called stress categorization. In order to perform the stress categorization to check results from FE models against the ASME Code stress limits, mainly from 3D solid FE models, several research works have been conducted. This work is included in this effort. First, a description of the ASME Code design criteria is presented. After that, a brief description of how the FEM can be used in pressure vessel design is showed. Several studies found in the literature on stress categorization for pressure vessel FE models are reviewed and commented. Then, the analyses done in this work are presented in which some typical nozzle to pressure vessel connections subjected to internal pressure and concentrated loads were modeled with solid finite elements. The results from linear elastic and limit load analyses are compared to each other and also with the results obtained by formulae for simple shell

  19. MO-F-CAMPUS-J-01: Effect of Iodine Contrast Agent Concentration On Cerebrovascular Dose for Synchrotron Radiation Microangiography Based On a Simple Mouse Head Model and a Voxel Mouse Head Phantom

    Energy Technology Data Exchange (ETDEWEB)

    Lin, H; Jing, J; Xie, C [Hefei University of Technology, Hefei (China); Lu, Y [Shanghai Jiao Tong University, Shanghai (China)

    2015-06-15

    Purpose: To find effective setting methods to mitigate the irradiation injure in synchrotron radiation microangiography(SRA) by Monte Carlo simulation. Methods: A mouse 1-D head model and a segmented voxel mouse head phantom were simulated by EGSnrc/Dosxyznrc code to investigate the dose enhancement effect of the iodine contrast agent irradiated by a monochromatic synchrotron radiation(SR) source. The influence of, like iodine concentration (IC), vessel width and depth, with and without skull layer protection and the various incident X ray energies, were simulated. The dose enhancement effect and the absolute dose based on the segmented voxel mouse head phantom were evaluated. Results: The dose enhancement ratio depends little on the irradiation depth, but strongly on the IC, which is linearly increases with IC. The skull layer protection cannot be ignored in SRA, the 700µm thick skull could decrease 10% of the dose. The incident X-ray energy can significantly affact the dose. E.g. compared to the dose of 33.2keV for 50mgI/ml, the 32.7keV dose decreases 38%, whereas the dose of 33.7 keV increases 69.2%, and the variation will strengthen more with enhanced IC. The segmented voxel mouse head phantom also showed that the average dose enhancement effect and the maximal voxel dose per photon depends little on the iodine voxel volume ratio, but strongly on IC. Conclusion: To decrease dose damage in SRA, the high-Z contrast agent should be used as little as possible, and try to avoid radiating locally the injected position immediately after the contrast agent injection. The fragile vessel containing iodine should avoid closely irradiating. Avoiding irradiating through the no or thin skull region, or appending thin equivalent material from outside to protect is also a better method. As long as SRA image quality is ensured, using incident X-ray energy as low as possible.

  20. 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.)

  1. 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.

  2. Corrosion Damage in Penetration Nozzle and Its Weldment of Reactor Pressure Vessel Head

    International Nuclear Information System (INIS)

    Lim, Yun Soo; Kim, Joung Soo; Kim, Hong Pyo; Hwang, Seong Sik; Yi, Young Sun; Kim, Dong Jin; Jung, Man Kyo

    2003-07-01

    The recent status on corrosion damage of reactor vessel head (RVH) penetration nozzles at primary water reactors (PWRs), including control rod drive mechanism (CRDM) and thermocouple nozzles, was investigated. The studies for primary water stress corrosion cracking (PWSCC) characteristics of Alloy 600 and Alloy 182/82 were reviewed and summarized in terms of the crack initiation and crack growth rate. The studies on the boric acid corrosion (BAC) of low alloy steels were also included in this report. PWSCC was found to be the main failure mechanism of RVH CRDM nozzles, which are constituted with Alloy 600 base metal and Alloy 182 weld filler materials. Alloy 600 and Alloy 182/82 are very susceptible to intergranular SCC in the PWR environments. The PWSCC crack initiation and growth features in the fusion zone of Alloy 182/82 were strongly dependant on solidification anisotropy during welding, test temperature, weld heat, mechanical loading, stress relief heat treatment, cold work and so on. BAC of low alloy steels is a wastage phenomenon due to general corrosion occurring on the over-all surface area of material. Systematic studies, concerned with structural integrity of RVH penetration nozzles as well as improvement of PWSCC resistance of nickel-based weld metals in the simulated PWR environment, are needed

  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. SCDAP/RELAP5 lower core plate model

    International Nuclear Information System (INIS)

    Coryell, E.W.; Griffin, F.P.

    1999-01-01

    The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. This report describes the justification, theory, implementation, and testing of a new modeling capability which will refine the analysis of the movement of molten material from the core region to the vessel lower head. As molten material moves from the core region through the core support structures it may encounter conditions which will cause it to freeze in the region of the lower core plate, delaying its arrival to the vessel head. The timing of this arrival is significant to reactor safety, because during the time span for material relocation to the lower head, the core may be experiencing steam-limited oxidation. The time at which hot material arrives in a coolant-filled lower vessel head, thereby significantly increasing the steam flow rate through the core region, becomes significant to the progression and timing of a severe accident. This report is a revision of a report INEEL/EXT-00707, entitled ''Preliminary Design Report for SCDAP/RELAP5 Lower Core Plate Model''

  5. 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

  6. Asymmetry of critical closing pressure following head injury

    OpenAIRE

    Kumar, A; Schmidt, E; Hiler, M; Smielewski, P; Pickard, J; Czosnyka, M

    2005-01-01

    Objective: Critical closing pressure (CCP) is the arterial pressure below which the vessels collapse. Hypothetically it is the sum of intracranial pressure (ICP) and vessel wall tension in the cerebral circulation. This study investigated transhemispherical asymmetry of CCP by studying its correlation with radiological findings on computed tomography (CT) scans in head injury patients.

  7. Combining endoscopes with PIV and digital holography for the study of vessel model mechanics

    International Nuclear Information System (INIS)

    Arévalo, Laura; Palero, Virginia; Andrés, Nieves; Arroyo, M P; Lobera, Julia

    2015-01-01

    In this work traditional fluid and solid mechanics measurement techniques have been combined with endoscopes for the study of blood vessel models’ mechanical properties. Endoscopes have been used as the imaging part of a high-speed PIV system to obtain the velocity field in a vessel model immersed in a container with a refractive index-matching liquid. In this way, we take advantage of the fact that the endoscope tip can be immersed in liquid. Endoscopes have also been used as the imaging and illuminating part of a digital holographic set-up for wall deformation measurement. The novelty of this work is that only one endoscope was used for illuminating and observing the vessel model, using the endoscope’s own illuminating system as the illumination source. The performance of endoscopes in different vessel models has been tested. The results of flow velocity and wall deformation in the different blood vessel models are presented. (paper)

  8. TMI-2 Vessel Investigation Project integration report

    International Nuclear Information System (INIS)

    Wolf, J.R.; Rempe, J.L.; Stickler, L.A.; Korth, G.E.; Diercks, D.R.; Neimark, L.A.; Akers, D.W.; Schuetz, B.K.; Shearer, T.L.; Chavez, S.A.; Thinnes, G.L.; Witt, R.J.; Corradini, M.L.; Kos, J.A.

    1994-03-01

    The Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) was an international effort that was sponsored by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. The primary objectives of the VIP were to extract and examine samples from the lower head and to evaluate the potential modes of failure and the margin of structural integrity that remained in the TMI-2 reactor vessel during the accident. This report presents a summary of the major findings and conclusions that were developed from research during the VIP. Results from the various elements of the project are integrated to form a cohesive understanding of the vessel's condition after the accident

  9. 3D Dynamic Modeling of the Head-Neck Complex for Fast Eye and Head Orientation Movements Research

    Directory of Open Access Journals (Sweden)

    Daniel A. Sierra

    2011-01-01

    Full Text Available A 3D dynamic computer model for the movement of the head-neck complex is presented. It incorporates anatomically correct information about the diverse elements forming the system. The skeleton is considered as a set of interconnected rigid 3D bodies following the Newton-Euler laws of movement. The muscles are modeled using Enderle's linear model, which shows equivalent dynamic characteristics to Loeb's virtual muscle model. The soft tissues, namely, the ligaments, intervertebral disks, and facet joints, are modeled considering their physiological roles and dynamics. In contrast with other head and neck models developed for safety research, the model is aimed to study the neural control of the complex during fast eye and head movements, such as saccades and gaze shifts. In particular, the time-optimal hypothesis and the feedback control ones are discussed.

  10. Theoretical modelling of physiologically stretched vessel in magnetisable stent assisted magnetic drug targeting application

    International Nuclear Information System (INIS)

    Mardinoglu, Adil; Cregg, P.J.; Murphy, Kieran; Curtin, Maurice; Prina-Mello, Adriele

    2011-01-01

    The magnetisable stent assisted magnetic targeted drug delivery system in a physiologically stretched vessel is considered theoretically. The changes in the mechanical behaviour of the vessel are analysed under the influence of mechanical forces generated by blood pressure. In this 2D mathematical model a ferromagnetic, coiled wire stent is implanted to aid collection of magnetic drug carrier particles in an elastic tube, which has similar mechanical properties to the blood vessel. A cyclic mechanical force is applied to the elastic tube to mimic the mechanical stress and strain of both the stent and vessel while in the body due to pulsatile blood circulation. The magnetic dipole-dipole and hydrodynamic interactions for multiple particles are included and agglomeration of particles is also modelled. The resulting collection efficiency of the mathematical model shows that the system performance can decrease by as much as 10% due to the effects of the pulsatile blood circulation. - Research highlights: →Theoretical modelling of magnetic drug targeting on a physiologically stretched stent-vessel system. →Cyclic mechanical force applied to mimic the mechanical stress and strain of both stent and vessel. →The magnetic dipole-dipole and hydrodynamic interactions for multiple particles is modelled. →Collection efficiency of the mathematical model is calculated for different physiological blood flow and magnetic field strength.

  11. Comparison of inverse modeling results with measured and interpolated hydraulic head data

    International Nuclear Information System (INIS)

    Jacobson, E.A.

    1986-12-01

    Inverse modeling of aquifers involves identification of effective parameters, such as transmissivities, based on hydraulic head data. The result of inverse modeling is a calibrated ground water flow model that reproduces the measured hydraulic head data as closely as is statistically possible. An inverse method that includes prior information about the parameters (i.e., kriged log transmissivity) was applied to the Avra Valley aquifer of southern Arizona using hydraulic heads obtained in three ways: measured at well locations, estimated at nodes by hand contouring, and estimated at nodes by kriging. Hand contouring yields only estimates of hydraulic head at node points, whereas kriging yields hydraulic head estimates at node points and their corresponding estimation errors. A comparison of the three inverse applications indicates the variations in the ground water flow model caused by the different treatments of the hydraulic head data. Estimates of hydraulic head computed by all three inverse models were more representative of the measured or interpolated hydraulic heads than those computed using the kriged estimates of log transmissivity. The large-scale trends in the estimates of log transmissivity determined by the three inverse models were generally similar except in the southern portion of the study area. The hydraulic head values and gradients produced by the three inverse models were similar in the interior of the study area, while the major differences between the inverse models occurred along the boundaries. 17 refs., 18 figs., 1 tab

  12. Reactor Vessel External Cooling for Corium Retention SULTAN Experimental Program and Modelling with CATHARE Code

    International Nuclear Information System (INIS)

    Rouge, S.; Dor, I.; Geffraye, G.

    1999-01-01

    In case of severe accident, a molten pool may form at the bottom of the lower head, and some pessimistic scenarios estimate that heat fluxes up to 1.5 MW/m 2 should be transferred through the vessel wall. An efficient, though completely passive, removal of heat flux during a long time is necessary to prevent total wall ablation, and a possible solution is to flood the cavity with water and establish boiling in natural convection. High heat exchanges are expected, especially if the system design (deflector along the vessel, riser...) emphasize water natural circulation, but are unfortunately limited by the critical heat flux phenomena (CHF). CHF data are very scarce in the adequate range of hydraulic and geometric parameters and are clearly dependent of the system effect in natural convection. The system effect can both modify flow velocity and two phase flow regimes, counter-current phenomena and flow static or dynamic instabilities. The SULTAN experimental program purpose was of two kinds, increasing CHF data for realistic situations, and improving the modeling of large 3D two phase flow circuits in natural convection. The CATHARE thermal-hydraulic code is used for interpreting the data and for extrapolation to real geometry. As a first step, a one-dimensional model is used. It is shown that some closure laws have to be improved. Reasonable predictions may be obtained but, for some test conditions, multi-dimensional effects such as recirculation appear to be dominant. Therefore the 3-dimensional module of CATHARE is also used to investigate these effects. This model well predicts qualitatively the existence and the development of a 2-phase layer along the heated wall as well as the existence of a recirculation zone. But modelling problems still require further development as part of a long term program for a better prediction of multi-dimensional two-phase flows

  13. Large-scale boiling experiments of the flooded cavity concept for in-vessel core retention

    International Nuclear Information System (INIS)

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

    1994-01-01

    This paper presents results of ex-vessel boiling experiments performed in the CYBL (CYlindrical BoiLing) facility. CYBL is a reactor-scale facility for confirmatory research of the flooded cavity concept for accident management. CYBL has a tank-within-a-tank design; the inner tank simulates the reactor vessel and the outer tank simulates the reactor cavity. Experiments with uniform and edge-peaked heat flux distributions up to 20 W/cm 2 across the vessel bottom were performed. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling is mainly due to the gravity head which results from flooding the sides 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 (ejection). The results suggest that under prototypic heat load and heat flux distributions, the flooded cavity in a passive pressurized water reactor like the AP-600 should be capable of cooling the reactor pressure vessel in the central region of the lower head that is addressed by these tests

  14. Thrust and torque characteristics based on a new cutter-head load model

    Science.gov (United States)

    Liu, Jianqin; Ren, Jiabao; Guo, Wei

    2015-07-01

    Full face rock tunnel boring machine(TBM) has been widely used in hard rock tunnels, however, there are few published theory about cutter-head design, and the design criteria of cutter-head under complex geological is not clear yet. To deal with the complex relationship among geological parameters, cutter parameters, and operating parameters during tunneling processes, a cutter-head load model is established by using CSM(Colorado school of mines) prediction model. Force distribution on cutter-head under a certain geology is calculated with the new established load model, and result shows that inner cutters bear more force than outer cutters, combining with disc cutters abrasion; a general principle of disc cutters' layout design is proposed. Within the model, the relationship among rock uniaxial compressive strength(UCS), penetration and thrust on cutter-head are analyzed, and the results shows that with increasing penetration, cutter thrust increases, but the growth rate slows and higher penetration makes lower special energy(SE). Finally, a fitting mathematical model of ZT(ratio of cutter-head torque and thrust) and penetration is established, and verified by TB880E, which can be used to direct how to set thrust and torque on cutter-head. When penetration is small, the cutter-head thrust is the main limiting factor in tunneling; when the penetration is large, cutter-head torque is the major limiting factor in tunneling. Based on the new cutter-head load model, thrust and torque characteristics of TBM further are researched and a new way for cutter-head layout design and TBM tunneling operations is proposed.

  15. Computed tomography depiction of small pediatric vessels with model-based iterative reconstruction

    Energy Technology Data Exchange (ETDEWEB)

    Koc, Gonca; Courtier, Jesse L.; Phelps, Andrew; Marcovici, Peter A.; MacKenzie, John D. [UCSF Benioff Children' s Hospital, Department of Radiology and Biomedical Imaging, San Francisco, CA (United States)

    2014-07-15

    Computed tomography (CT) is extremely important in characterizing blood vessel anatomy and vascular lesions in children. Recent advances in CT reconstruction technology hold promise for improved image quality and also reductions in radiation dose. This report evaluates potential improvements in image quality for the depiction of small pediatric vessels with model-based iterative reconstruction (Veo trademark), a technique developed to improve image quality and reduce noise. To evaluate Veo trademark as an improved method when compared to adaptive statistical iterative reconstruction (ASIR trademark) for the depiction of small vessels on pediatric CT. Seventeen patients (mean age: 3.4 years, range: 2 days to 10.0 years; 6 girls, 11 boys) underwent contrast-enhanced CT examinations of the chest and abdomen in this HIPAA compliant and institutional review board approved study. Raw data were reconstructed into separate image datasets using Veo trademark and ASIR trademark algorithms (GE Medical Systems, Milwaukee, WI). Four blinded radiologists subjectively evaluated image quality. The pulmonary, hepatic, splenic and renal arteries were evaluated for the length and number of branches depicted. Datasets were compared with parametric and non-parametric statistical tests. Readers stated a preference for Veo trademark over ASIR trademark images when subjectively evaluating image quality criteria for vessel definition, image noise and resolution of small anatomical structures. The mean image noise in the aorta and fat was significantly less for Veo trademark vs. ASIR trademark reconstructed images. Quantitative measurements of mean vessel lengths and number of branches vessels delineated were significantly different for Veo trademark and ASIR trademark images. Veo trademark consistently showed more of the vessel anatomy: longer vessel length and more branching vessels. When compared to the more established adaptive statistical iterative reconstruction algorithm, model

  16. Computed tomography depiction of small pediatric vessels with model-based iterative reconstruction

    International Nuclear Information System (INIS)

    Koc, Gonca; Courtier, Jesse L.; Phelps, Andrew; Marcovici, Peter A.; MacKenzie, John D.

    2014-01-01

    Computed tomography (CT) is extremely important in characterizing blood vessel anatomy and vascular lesions in children. Recent advances in CT reconstruction technology hold promise for improved image quality and also reductions in radiation dose. This report evaluates potential improvements in image quality for the depiction of small pediatric vessels with model-based iterative reconstruction (Veo trademark), a technique developed to improve image quality and reduce noise. To evaluate Veo trademark as an improved method when compared to adaptive statistical iterative reconstruction (ASIR trademark) for the depiction of small vessels on pediatric CT. Seventeen patients (mean age: 3.4 years, range: 2 days to 10.0 years; 6 girls, 11 boys) underwent contrast-enhanced CT examinations of the chest and abdomen in this HIPAA compliant and institutional review board approved study. Raw data were reconstructed into separate image datasets using Veo trademark and ASIR trademark algorithms (GE Medical Systems, Milwaukee, WI). Four blinded radiologists subjectively evaluated image quality. The pulmonary, hepatic, splenic and renal arteries were evaluated for the length and number of branches depicted. Datasets were compared with parametric and non-parametric statistical tests. Readers stated a preference for Veo trademark over ASIR trademark images when subjectively evaluating image quality criteria for vessel definition, image noise and resolution of small anatomical structures. The mean image noise in the aorta and fat was significantly less for Veo trademark vs. ASIR trademark reconstructed images. Quantitative measurements of mean vessel lengths and number of branches vessels delineated were significantly different for Veo trademark and ASIR trademark images. Veo trademark consistently showed more of the vessel anatomy: longer vessel length and more branching vessels. When compared to the more established adaptive statistical iterative reconstruction algorithm, model

  17. Comparison study of different head model structures with homogeneous/inhomogeneous conductivity

    International Nuclear Information System (INIS)

    Wen, P.; Li, Y.

    2001-01-01

    Most of the human head models used in dipole localisation research, which have been reported in the literature to date, assume a simplified cranial structure wherein the head is modelled as a set of distinct homogenous tissue compartments. The inherent inhomogeneity of the tissues has so far been ignored in these models due to the difficulties involved in obtaining the conductivity characteristics with sufficiently high enough spatial resolution throughout the head. A technique for developing an inhomogeneous head model based on the generation of pseudo-conductivity values from the existing but sparse conductivity values is proposed in this paper. Comparative studies are conducted on different model structures and different mechanisms for generating the pseudo conductivities. An evaluation of the results of these studies as reported in this paper, shows that contrary to current simplifying assumptions, tissue inhomogeneity has a major influence on the computation of electrical potential distributions in the head. Brain electrical activity is spatially distributed in three dimensions in the head and evolves with time. Electroencephalography (EEG) is a widely used noninvasive technique which measures the potential distribution on the scalp caused by the brain electrical activity. A number of interesting correlations between features of the recorded EEG waveforms and various aspects of attention memory and linguistic tAS/Ks have been discovered. These correlations are estimated by comparing, for a given brain function, the recorded EEGs against the scalp potentials obtained from the computation of an electric field model of the head. The accuracy of these estimates depends not only on such factors as EEG measured errors but also, more importantly, on how closely the head model approximates the physiological head. This has spurred interest in the use of a more realistic head geometry with more accurate conductivity values which would use the detailed anatomical

  18. Safety of nuclear pressure vessels and its regulatory aspects in France

    Energy Technology Data Exchange (ETDEWEB)

    de Torquat, G; Queniart, D; Barrachin, B; Roche, R

    1979-01-01

    Having outlined the basic French regulations governing the safety of both pressure vessels and also of nuclear installations in general the particular safety regulations covering prestressed concrete vessels for nuclear reactors are considered. The regulations now being prepared to cover heat transfer systems of water reactors are detailed under sections headed; general provisions, sizing, and construction.

  19. Development of severe accident analysis code - Development of a finite element code for lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Lee, Choong Ho; Choi, Tae Hoon; Kim, Hyun Sup; Kim, Se Ho; Kang, Woo Jong; Seo, Chong Kwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-08-01

    The study concerns the development of analysis models and computer codes for lower head failure analysis when a severe accident occurs in a nuclear reactor system. Although the lower head failure modes consists of several failure modes, the study this year was focused on the global rupture with the collapse pressure and mode by limit analysis and elastic deformation. The behavior of molten core causes elevation of temperature in the reactor vessel wall and deterioration of load-carrying capacity of a reactor vessel. The behavior of molten core and the heat transfer modes were, therefore, postulated in several types and the temperature distributions according to the assumed heat flux modes were calculated. The collapse pressure of a nuclear reactor lower head decreases rapidly with elevation of temperature as time passes. The calculation shows the safety of a nuclear reactor is enhanced with the lager collapse pressure when the hot spot is located far from the pole. 42 refs., 2 tabs., 31 figs. (author)

  20. Analysis of the consequences of the anomaly in the Flamanville EPR reactor pressure vessel head domes on their serviceability. Report to the Advisory Committee of Experts for Nuclear Pressure Equipment. Public version. Session of 26 and 27 June 2017

    International Nuclear Information System (INIS)

    CATTEAU, R.; HERVIOU, K.

    2017-06-01

    substantiating the fact that the material of the Flamanville EPR reactor pressure vessel head closure and bottom head domes is ductile and tough enough to deal with the operating conditions of this equipment. This file more particularly draws on the results of the mechanical tests and concludes that the domes are serviceable. In its letter in reference, ASN informed Areva NP that it considered that the technical qualification requirement of the ESPN order in reference was not met for the domes, because the heterogeneity risk had been poorly assessed and the characteristics of the material were not as expected. Areva NP thus envisages sending ASN a commissioning authorisation application for the Flamanville EPR reactor pressure vessel, even though it has not met all the regulatory requirements, pursuant to article 93 of the ESPN order in reference. This report is a part of the advance technical examination of this authorisation application. In its letter in reference, ASN informed Areva NP that such an application needed to be substantiated with regard to the advantages and drawbacks of alternative solutions, notably repair of the reactor pressure vessel and replacement of the closure head. Areva NP considers that procurement of a new closure head and replacement of the existing one, an operation that has already been carried out on several reactors, would take at least 75 months. Areva NP and EDF also examined the possibility of repairing the reactor pressure vessel bottom head and consider that the consequences would be disproportionate in terms of cost, lead-time and consequences for the EPR reactor model and the nuclear reactor system. Repair would entail extracting the reactor pressure vessel from its cavity, replacing its bottom head, reinstalling it and rebuilding a part of the surrounding civil engineering structures. These operations are estimated to take 86 months. These various aspects, which are not examined within the framework of this report, are detailed in

  1. Modeling transient streaming potentials in falling-head permeameter tests.

    Science.gov (United States)

    Malama, Bwalya; Revil, André

    2014-01-01

    We present transient streaming potential data collected during falling-head permeameter tests performed on samples of two sands with different physical and chemical properties. The objective of the work is to estimate hydraulic conductivity (K) and the electrokinetic coupling coefficient (Cl ) of the sand samples. A semi-empirical model based on the falling-head permeameter flow model and electrokinetic coupling is used to analyze the streaming potential data and to estimate K and Cl . The values of K estimated from head data are used to validate the streaming potential method. Estimates of K from streaming potential data closely match those obtained from the associated head data, with less than 10% deviation. The electrokinetic coupling coefficient was estimated from streaming potential vs. (1) time and (2) head data for both sands. The results indicate that, within limits of experimental error, the values of Cl estimated by the two methods are essentially the same. The results of this work demonstrate that a temporal record of the streaming potential response in falling-head permeameter tests can be used to estimate both K and Cl . They further indicate the potential for using transient streaming potential data as a proxy for hydraulic head in hydrogeology applications. © 2013, National Ground Water Association.

  2. Continuum mathematical modelling of pathological growth of blood vessels

    Science.gov (United States)

    Stadnik, N. E.; Dats, E. P.

    2018-04-01

    The present study is devoted to the mathematical modelling of a human blood vessel pathological growth. The vessels are simulated as the thin-walled circular tube. The boundary value problem of the surface growth of an elastic thin-walled cylinder is solved. The analytical solution is obtained in terms of velocities of stress strain state parameters. The condition of thinness allows us to study finite displacements of cylinder surfaces by means of infinitesimal deformations. The stress-strain state characteristics, which depend on the mechanical parameters of the biological processes, are numerically computed and graphically analysed.

  3. Performance experiments on the in-vessel core catcher during severe accidents

    International Nuclear Information System (INIS)

    Kang, Kyoung Ho; Park, Rae Joon; Cho, Young Rho; Kim, Sang Baik

    2004-01-01

    A US-Korean International Nuclear Energy Research Initiative (INERI) project has been initiated by the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korean Atomic Energy Research Institute (KAERI) to determine if IVR is feasible for high power reactors up to 1500 MWe by investigating the performance of enhanced ERVC and in-vessel core catcher. This program is initially focusing on the Korean Advanced Power Reactor 1400 MWe (APR1400) design. As for the enhancement of the coolability through the ERVC, boiling tests are conducted by using appropriate coating material on the vessel outer surface to promote downward facing boiling and selecting an improved vessel/insulation design to facilitate water flow and steam venting through the insulation in this program. Another approach for successful IVR are investigated by applying the in-vessel core catcher to provide an 'engineered gap' between the relocated core materials and the water-filled reactor vessel and a preliminary design for an in-vessel core catcher was developed during the first year of this program. Feasibility experiments using the LAVA facility, named LAVA-GAP experiments, are in progress to investigate the core catcher performance based on the conceptual design of the in-vessel core catcher proposed in this INERI project. The experiments were performed using 60kg of Al 2 O 3 thermite melt as a core material simulant with a 1/8 linear scale mock-up of the reactor vessel lower plenum. The hemispherical in-vessel core catcher was installed inside the lower head vessel maintaining a uniform gap of 10mm from the inner surface of the lower head vessel. Two types of the core catchers were used in these experiments. The first one was a single layered in-vessel core catcher without internal coating and the second one was a two layered in-vessel core catcher with an internal coating of 0.5mm-thick ZrO 2 via the plasma

  4. A model for ultrasound contrast agent in a phantom vessel

    KAUST Repository

    Qamar, Adnan; Samtaney, Ravi

    2014-01-01

    A theoretical framework to model the dynamics of Ultrasound Contrast Agent (UCA) inside a phantom vessel is presented. The model is derived from the reduced Navier-Stokes equation and is coupled with the evolving flow field solution inside

  5. Development of simplified 1D and 2D models for studying a PWR lower head failure under severe accident conditions

    International Nuclear Information System (INIS)

    Koundy, V.; Dupas, J.; Bonneville, H.; Cormeau, I.

    2005-01-01

    In the study of severe accidents of nuclear pressurized water reactors, the scenarios that describe the relocation of significant quantities of liquid corium at the bottom of the lower head are investigated from the mechanical point of view. In these scenarios, the risk of a breach and the possibility of a large quantity of corium being released from the lower head exist. This may lead to direct heating of the containment or outer vessel steam explosion. These issues are important due to their early containment failure potential. Since the TMI-2 accident, many theoretical and experimental investigations, relating to lower head mechanical behaviour under severe thermo-mechanical loading in the event of a core meltdown accident have been performed. IRSN participated actively in the one-fifth scale USNRC/SNL LHF and OECD LHF (OLHF) programs. Within the framework of these programs, two simplified models were developed by IRSN: the first is a simplified 1D approach based on the theory of pressurized spherical shells and the second is a simplified 2D model based on the theory of shells of revolution under symmetric loading. The mathematical formulation of both models and the creep constitutive equations used are presented in detail in this paper. The corresponding models were used to interpret some of the OLHF program experiments and the calculation results were quite consistent with the experimental data. The two simplified models have been used to simulate the thermo-mechanical behaviour of a 900 MWe pressurized water reactor lower head under severe accident conditions leading to failure. The average transient heat flux produced by the corium relocated at the bottom of the lower head has been determined using the IRSN HARAR code. Two different methods, both taking into account the ablation of the internal surface, are used to determine the temperature profiles across the lower head wall and their effect on the time to failure is discussed. Using these simplified models

  6. Parametric model to estimate containment loads following an ex-vessel steam spike

    International Nuclear Information System (INIS)

    Lopez, R.; Hernandez, J.; Huerta, A.

    1998-01-01

    This paper describes the use of a relatively simple parametric model to estimate containment loads following an ex-vessel steam spike. The study was motivated because several PSAs have identified containment loads accompanying reactor vessel failures as a major contributor to early containment failure. The paper includes a detailed description of the simple but physically sound parametric model which was adopted to estimate containment loads following a steam spike into the reactor cavity. (author)

  7. Improved Wave-vessel Transfer Functions by Uncertainty Modelling

    DEFF Research Database (Denmark)

    Nielsen, Ulrik Dam; Fønss Bach, Kasper; Iseki, Toshio

    2016-01-01

    This paper deals with uncertainty modelling of wave-vessel transfer functions used to calculate or predict wave-induced responses of a ship in a seaway. Although transfer functions, in theory, can be calculated to exactly reflect the behaviour of the ship when exposed to waves, uncertainty in inp...

  8. 33 CFR 164.35 - Equipment: All vessels.

    Science.gov (United States)

    2010-07-01

    ... to alter course 90 degrees with maximum rudder angle and constant power settings, for either full and... communication for relaying headings to the emergency steering station. Also, each vessel of 500 gross tons and over and constructed on or after June 9, 1995 must be provided with arrangements for supplying visual...

  9. Sensitivity analyses on in-vessel hydrogen generation for KNGR

    International Nuclear Information System (INIS)

    Kim, See Darl; Park, S.Y.; Park, S.H.; Park, J.H.

    2001-03-01

    Sensitivity analyses for the in-vessel hydrogen generation, using the MELCOR program, are described in this report for the Korean Next Generation Reactor. The typical accident sequences of a station blackout and a large LOCA scenario are selected. A lower head failure model, a Zircaloy oxidation reaction model and a B 4 C reaction model are considered for the sensitivity parameters. As for the base case, 1273.15K for a failure temperature of the penetrations or the lower head, an Urbanic-Heidrich correlation for the Zircaloy oxidation reaction model and the B 4 C reaction model are used. Case 1 used 1650K as the failure temperature for the penetrations and Case 2 considered creep rupture instead of penetration failure. Case 3 used a MATPRO-EG and G correlation for the Zircaloy oxidation reaction model and Case 4 turned off the B 4 C reaction model. The results of the studies are summarized below : (1) When the penetration failure temperature is higher, or the creep rupture failure model is considered, the amount of hydrogen increases for two sequences. (2) When the MATPRO-EG and G correlation for a Zircaloy oxidation reaction is considered, the amount of hydrogen is less than the Urbanic-Heidrich correlation (Base case) for both scenarios. (3) When the B 4 C reaction model turns off, the amount of hydrogen decreases for two sequences

  10. Establishment of an animal model of mice with radiation- injured soft tissue blood vessels

    International Nuclear Information System (INIS)

    Wang Daiyou; Yu Dahai; Wu Jiaxiao; Wei Shanliang; Wen Yuming

    2004-01-01

    Objective: The aim of this study was to establish an animal model of mice with radiation-injured soft tissue blood vessels. Methods: Forty male mice were irradiated with 30 Gy on the right leg. After the irradiation was finished each of the 40 male mice was tested with angiography, and its muscle tissues on the bilateral legs were examined with vessel staining assay and electron microscopy. Results: The results showed that the number of vessels on the right leg was less than that on the left leg, the microvessel density, average diameter and average sectional area of the right leg were all lower than those of the left, and the configuration and ultra-structure of vessels were also different between both sides of legs. Conclusion: In the study authors successfully established an animal model of mice with radiation-injured soft tissue blood vessels

  11. Simplified realistic human head model for simulating Tumor Treating Fields (TTFields).

    Science.gov (United States)

    Wenger, Cornelia; Bomzon, Ze'ev; Salvador, Ricardo; Basser, Peter J; Miranda, Pedro C

    2016-08-01

    Tumor Treating Fields (TTFields) are alternating electric fields in the intermediate frequency range (100-300 kHz) of low-intensity (1-3 V/cm). TTFields are an anti-mitotic treatment against solid tumors, which are approved for Glioblastoma Multiforme (GBM) patients. These electric fields are induced non-invasively by transducer arrays placed directly on the patient's scalp. Cell culture experiments showed that treatment efficacy is dependent on the induced field intensity. In clinical practice, a software called NovoTalTM uses head measurements to estimate the optimal array placement to maximize the electric field delivery to the tumor. Computational studies predict an increase in the tumor's electric field strength when adapting transducer arrays to its location. Ideally, a personalized head model could be created for each patient, to calculate the electric field distribution for the specific situation. Thus, the optimal transducer layout could be inferred from field calculation rather than distance measurements. Nonetheless, creating realistic head models of patients is time-consuming and often needs user interaction, because automated image segmentation is prone to failure. This study presents a first approach to creating simplified head models consisting of convex hulls of the tissue layers. The model is able to account for anisotropic conductivity in the cortical tissues by using a tensor representation estimated from Diffusion Tensor Imaging. The induced electric field distribution is compared in the simplified and realistic head models. The average field intensities in the brain and tumor are generally slightly higher in the realistic head model, with a maximal ratio of 114% for a simplified model with reasonable layer thicknesses. Thus, the present pipeline is a fast and efficient means towards personalized head models with less complexity involved in characterizing tissue interfaces, while enabling accurate predictions of electric field distribution.

  12. Development of a master model concept for DEMO vacuum vessel

    International Nuclear Information System (INIS)

    Mozzillo, Rocco; Marzullo, Domenico; Tarallo, Andrea; Bachmann, Christian; Di Gironimo, Giuseppe

    2016-01-01

    Highlights: • The present work concerns the development of a first master concept model for DEMO vacuum vessel. • A parametric-associative CAD master model concept of a DEMO VV sector has been developed in accordance with DEMO design guidelines. • A proper CAD design methodology has been implemented in view of the later FEM analyses based on “shell elements”. - Abstract: This paper describes the development of a master model concept of the DEMO vacuum vessel (VV) conducted within the framework of the EUROfusion Consortium. Starting from the VV space envelope defined in the DEMO baseline design 2014, the layout of the VV structure was preliminarily defined according to the design criteria provided in RCC-MRx. A surface modelling technique was adopted and efficiently linked to the finite element (FE) code to simplify future FE analyses. In view of possible changes to shape and structure during the conceptual design activities, a parametric design approach allows incorporating modifications to the model efficiently.

  13. Development of a master model concept for DEMO vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Mozzillo, Rocco; Marzullo, Domenico; Tarallo, Andrea [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Bachmann, Christian [EUROfusion PMU, Boltzmannstraße 2, 85748 Garching (Germany); Di Gironimo, Giuseppe, E-mail: peppe.digironimo@gmail.com [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy)

    2016-11-15

    Highlights: • The present work concerns the development of a first master concept model for DEMO vacuum vessel. • A parametric-associative CAD master model concept of a DEMO VV sector has been developed in accordance with DEMO design guidelines. • A proper CAD design methodology has been implemented in view of the later FEM analyses based on “shell elements”. - Abstract: This paper describes the development of a master model concept of the DEMO vacuum vessel (VV) conducted within the framework of the EUROfusion Consortium. Starting from the VV space envelope defined in the DEMO baseline design 2014, the layout of the VV structure was preliminarily defined according to the design criteria provided in RCC-MRx. A surface modelling technique was adopted and efficiently linked to the finite element (FE) code to simplify future FE analyses. In view of possible changes to shape and structure during the conceptual design activities, a parametric design approach allows incorporating modifications to the model efficiently.

  14. 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

  15. Control Rod Drive Mechanism Installed in the Internal of Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, M. H.; Choi, S.; Park, J. S.; Lee, J. S.; Kim, D. O.; Hur, N. S.; Hur, H.; Yu, J. Y

    2008-09-15

    This report describes the review results and important technologies related to the in-vessel type control rod drive mechanism. Generally, most of the CRDMs used in the PWR are attached outside of the reactor pressure vessel, and the pernetration of the vessel head can not avoid. However, in-vessel type CRDMs, which are installed inside the reactor vessel, can eliminate the possibility of rod ejection accidents and the penetration of the vessel head, and provide a compact design of the reactor vessel and containment. There are two kinds of in-vessel type CRDM concerning the driving force-driven by a driving motor and by a hydraulic force. Motor driven CRDMs have been mainly investigated in Japan(MRX, IMR, DRX, next generation BWR etc.), and developed the key components such as a canned motor, an integrated rod position indicator, a separating ball-nut and a ball bearing that can operate under the water conditions of a high temperature and pressure. The concept of hydraulically driven CRDMs have been first reported by KWU and Siemens for KWU 200 reactor, and Argentina(CAREM) and China(NHR-5, NHR-200) have been developed the internal CRDM with the piston and cylinder of slightly different geometries. These systems are driven by the hydraulic force which is produced by pumps outside of the reactor vessel and transmitted through a pipe penetrating the reactor vessel, and needs complicated control and piping systems including pumps, valves and pipes etc.. IRIS has been recently decided the internal CRDMs as the reference design, and an analytical and experimental investigations of the hydraulic drive concept are performed by POLIMI in Italy. Also, a small French company, MP98 has been developed a new type of control rods, called 'liquid control rods', where reactivity is controlled by the movement of a liquid absorber in a manometer type device.

  16. Repairing method for shroud in reactor pressure vessel

    International Nuclear Information System (INIS)

    Watanabe, Yusuke.

    1996-01-01

    The present invention provides a method of repairing a shroud disposed in a pressure vessel of a BWR type reactor. Namely, a baffle plate is disposed on the outer surface of the lower portion of the shroud supported by a shroud support of the pressure vessel. The baffle plate is connected with a lug for securing a shroud head bolt disposed on the outer surface of an upper portion of the shroud by reinforcing members. With such a constitution, when crackings are caused in the shroud, the development of the crackings can be prevented without losing the function of securing the shroud head bolt. Further, if a material having thermal expansion coefficient lower than that of austenite stainless steel is used for the material of the reinforcing member, clamping load to be applied upon attaching the auxiliary member can be reduced. As a result, operation for the attachment is facilitated. (I.S.)

  17. Simulant melt experiments on performance of the in-vessel core catcher

    International Nuclear Information System (INIS)

    Kyoung-Ho Kang; Rae-Joon Park; Sang-Baik Kim; Suh, K.Y.; Cheung, F.B.; Rempe, J.L.

    2005-01-01

    Full text of publication follows: LAVA-GAP experiments are in progress to investigate the performance of the in-vessel core catcher using alumina melt as a corium simulant. The hemispherical in-vessel core catcher made of carbon steel was installed inside the lower head vessel with uniform gap of 5 mm or 10 mm to the inner surface of the lower head vessel. As a performance test of the in-vessel core catcher, the effects of base steel and internal coating materials and gap thickness between the core catcher and the lower head vessel were examined in this study. In the LAVA-GAP-2 and LAVA-GAP-3 tests, the base steel was carbon steel and the gap thickness was 10 mm. On the other hand, in the LAVA-GAP-4 and LAVA-GAP-5 tests, the base steel was stainless steel and the gap thickness was 5 mm. Actual composition of the coating material for the LAVA-GAP-4 test was 92% of ZrO 2 - 8% of Y 2 O 3 including 95% of Ni - 5% of Al bond coat same as the LAVA-GAP-3 test. In these tests, the thickness of ZrO 2 internal coating was 0.5 mm. To examine the effects of the coating material, in-vessel core catcher with a 0.6 mm-thick ZrO 2 coating without bond coat was used in the LAVA-GAP-5 test. This report summarizes the experimental results and the post metallurgical inspection results of the LAVA-GAP-4 and LAVA-GAP- 5 tests. In the LAVA-GAP-4 and LAVA-GAP-5 tests, the core catcher was failed and it was stuck to the inner surface of the lower head vessel. LAVA-GAP-4 and LAVA-GAP-5 test results imply that 5 mm thick gap is rather small for sufficient water ingression and steam venting through the gap. In case of small gap size, water is boiled off and steam increases pressure inside the gap and so water can not ingress into the gap at the initial heat up stage. Metallurgical inspections on the test specimens indicate that the internal coating layer might melt totally and dispersed in the base steel and the solidified iron melt and so the detection frequencies of Zr and O are trivial all

  18. A unified model of heading and path perception in primate MSTd.

    Directory of Open Access Journals (Sweden)

    Oliver W Layton

    2014-02-01

    Full Text Available Self-motion, steering, and obstacle avoidance during navigation in the real world require humans to travel along curved paths. Many perceptual models have been proposed that focus on heading, which specifies the direction of travel along straight paths, but not on path curvature, which humans accurately perceive and is critical to everyday locomotion. In primates, including humans, dorsal medial superior temporal area (MSTd has been implicated in heading perception. However, the majority of MSTd neurons respond optimally to spiral patterns, rather than to the radial expansion patterns associated with heading. No existing theory of curved path perception explains the neural mechanisms by which humans accurately assess path and no functional role for spiral-tuned cells has yet been proposed. Here we present a computational model that demonstrates how the continuum of observed cells (radial to circular in MSTd can simultaneously code curvature and heading across the neural population. Curvature is encoded through the spirality of the most active cell, and heading is encoded through the visuotopic location of the center of the most active cell's receptive field. Model curvature and heading errors fit those made by humans. Our model challenges the view that the function of MSTd is heading estimation, based on our analysis we claim that it is primarily concerned with trajectory estimation and the simultaneous representation of both curvature and heading. In our model, temporal dynamics afford time-history in the neural representation of optic flow, which may modulate its structure. This has far-reaching implications for the interpretation of studies that assume that optic flow is, and should be, represented as an instantaneous vector field. Our results suggest that spiral motion patterns that emerge in spatio-temporal optic flow are essential for guiding self-motion along complex trajectories, and that cells in MSTd are specifically tuned to extract

  19. A Unified Model of Heading and Path Perception in Primate MSTd

    Science.gov (United States)

    Layton, Oliver W.; Browning, N. Andrew

    2014-01-01

    Self-motion, steering, and obstacle avoidance during navigation in the real world require humans to travel along curved paths. Many perceptual models have been proposed that focus on heading, which specifies the direction of travel along straight paths, but not on path curvature, which humans accurately perceive and is critical to everyday locomotion. In primates, including humans, dorsal medial superior temporal area (MSTd) has been implicated in heading perception. However, the majority of MSTd neurons respond optimally to spiral patterns, rather than to the radial expansion patterns associated with heading. No existing theory of curved path perception explains the neural mechanisms by which humans accurately assess path and no functional role for spiral-tuned cells has yet been proposed. Here we present a computational model that demonstrates how the continuum of observed cells (radial to circular) in MSTd can simultaneously code curvature and heading across the neural population. Curvature is encoded through the spirality of the most active cell, and heading is encoded through the visuotopic location of the center of the most active cell's receptive field. Model curvature and heading errors fit those made by humans. Our model challenges the view that the function of MSTd is heading estimation, based on our analysis we claim that it is primarily concerned with trajectory estimation and the simultaneous representation of both curvature and heading. In our model, temporal dynamics afford time-history in the neural representation of optic flow, which may modulate its structure. This has far-reaching implications for the interpretation of studies that assume that optic flow is, and should be, represented as an instantaneous vector field. Our results suggest that spiral motion patterns that emerge in spatio-temporal optic flow are essential for guiding self-motion along complex trajectories, and that cells in MSTd are specifically tuned to extract complex trajectory

  20. LANL Robotic Vessel Scanning

    Energy Technology Data Exchange (ETDEWEB)

    Webber, Nels W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-25

    Los Alamos National Laboratory in J-1 DARHT Operations Group uses 6ft spherical vessels to contain hazardous materials produced in a hydrodynamic experiment. These contaminated vessels must be analyzed by means of a worker entering the vessel to locate, measure, and document every penetration mark on the vessel. If the worker can be replaced by a highly automated robotic system with a high precision scanner, it will eliminate the risks to the worker and provide management with an accurate 3D model of the vessel presenting the existing damage with the flexibility to manipulate the model for better and more in-depth assessment.The project was successful in meeting the primary goal of installing an automated system which scanned a 6ft vessel with an elapsed time of 45 minutes. This robotic system reduces the total time for the original scope of work by 75 minutes and results in excellent data accumulation and transmission to the 3D model imaging program.

  1. TMI-2 Vessel Investigation Project (VIP) Metallurgical Program

    International Nuclear Information System (INIS)

    Diercks, D.R.; Neimark, L.A.

    1991-01-01

    The Three Mile Island Unite 2 (TMI-2) Vessel Investigation Project Metallurgical Program is a part of the international TMI-2 Vessel Investigation Project being conducted jointly by the U.S. Nuclear Regulatory Commission and the Organization for Economic Cooperation and Development. The objectives of the metallurgical program are to deduce the temperatures of, determine the mechanical properties of, and assess the integrity of the TMI-2 lower head during the loss-of-coolant accident. Fifteen samples have been removed from the lower head and are being examined. In addition, archive material from the lower head of the Midland nuclear reactor has been procured for conducting supplemental metallurgical evaluations and mechanical property determinations. Evaluations of the microstructure and mechanical properties of the as-received archive material have been completed, and a series of heat treatment experiments has been conducted to develop standard microstructures to be compared with those present in the TMI-2 samples. Results have been obtained from examinations of two of the fifteen TMI-2 lower head samples. These results indicate that one of these two samples, which contained cracks in the weld cladding extending ∼3 mm into the underlying base metal, apparently reached temperatures on the order of 1000 to 1100C during the accident. A preliminary examination of the core debris deposited on this sample has been performed. The other sample, from an area away from the region of core relocation, did not exceed 727C during the accident

  2. Dynamic Positioning Capability Analysis for Marine Vessels Based on A DPCap Polar Plot Program

    Science.gov (United States)

    Wang, Lei; Yang, Jian-min; Xu, Sheng-wen

    2018-03-01

    Dynamic positioning capability (DPCap) analysis is essential in the selection of thrusters, in their configuration, and during preliminary investigation of the positioning ability of a newly designed vessel dynamic positioning system. DPCap analysis can help determine the maximum environmental forces, in which the DP system can counteract in given headings. The accuracy of the DPCap analysis is determined by the precise estimation of the environmental forces as well as the effectiveness of the thrust allocation logic. This paper is dedicated to developing an effective and efficient software program for the DPCap analysis for marine vessels. Estimation of the environmental forces can be obtained by model tests, hydrodynamic computation and empirical formulas. A quadratic programming method is adopted to allocate the total thrust on every thruster of the vessel. A detailed description of the thrust allocation logic of the software program is given. The effectiveness of the new program DPCap Polar Plot (DPCPP) was validated by a DPCap analysis for a supply vessel. The present study indicates that the developed program can be used in the DPCap analysis for marine vessels. Moreover, DPCap analysis considering the thruster failure mode might give guidance to the designers of vessels whose thrusters need to be safer.

  3. Structural analysis and evaluation for the design of pressure vessel

    International Nuclear Information System (INIS)

    Arai, K.; Uragami, K.; Funada, T.; Baba, K.; Kira, T.

    1977-01-01

    For the design of pressure vessel, the detailed structural analysis such as the fatigue analysis under operating conditions is required by ASME Code or Japanese regulation. Accordingly, it should be verified by the analysis that the design of the pressure vessel is in compliance with the stress limitation defined in the Code or the regulation. However, it was apparent that the analysis is very complicated and takes a lot of time to evaluate in accordance with the Code requirements. Thereupon we developed the computer program by which we can perform the stress analysis with correctness and comparatively in a short period of design work reflecting the calculation results on detailed drawings to be used for fabrication. The computer program is controlled in combination with the system of the design work and out put list of the program can be directly used for the stress analysis report which is issued to customers. In addition to the above computer program, we developed the specific three dimensional finite element computer program to make sure of the structural integrity of the vessel head and flanges which are most complex for the analysis compared with the stress distribution measured by strain gauges on the vessel head and flange. Besides the structural analysis, the fracture mechanics analysis for the purpose of preventing the pressure vessel from the brittle fracture during heat-up and cool-down operation is also important and thereby we showed herein that the pressure vessel is in safety against the brittle fracture for the specified operating conditions. As a result of the above-mentioned analysis, the pressure vessel is designed with safety from the stand-points of the structural intensity and the fracture mechanics. (auth.)

  4. Reactor pressure vessel failure probability following through-wall cracks due to pressurized thermal shock events

    International Nuclear Information System (INIS)

    Simonen, F.A.; Garnich, M.R.; Simonen, E.P.; Bian, S.H.; Nomura, K.K.; Anderson, W.E.; Pedersen, L.T.

    1986-04-01

    A fracture mechanics model was developed at the Pacific Northwest Laboratory (PNL) to predict the behavior of a reactor pressure vessel following a through-wall crack that occurs during a pressurized thermal shock (PTS) event. This study, which contributed to a US Nuclear Regulatory Commission (NRC) program to study PTS risk, was coordinated with the Integrated Pressurized Thermal Shock (IPTS) Program at Oak Ridge National Laboratory (ORNL). The PNL fracture mechanics model uses the critical transients and probabilities of through-wall cracks from the IPTS Program. The PNL model predicts the arrest, reinitiation, and direction of crack growth for a postulated through-wall crack and thereby predicts the mode of vessel failure. A Monte-Carlo type of computer code was written to predict the probabilities of the alternative failure modes. This code treats the fracture mechanics properties of the various welds and plates of a vessel as random variables. Plant-specific calculations were performed for the Oconee-1, Calvert Cliffs-1, and H.B. Robinson-2 reactor pressure vessels for the conditions of postulated transients. The model predicted that 50% or more of the through-wall axial cracks will turn to follow a circumferential weld. The predicted failure mode is a complete circumferential fracture of the vessel, which results in a potential vertically directed missile consisting of the upper head assembly. Missile arrest calculations for the three nuclear plants predict that such vertical missiles, as well as all potential horizontally directed fragmentation type missiles, will be confined to the vessel enclosre cavity. The PNL failure mode model is recommended for use in future evaluations of other plants, to determine the failure modes that are most probable for postulated PTS events

  5. A Novel Through Capacity Model for One-way Channel Based on Characteristics of the Vessel Traffic Flow

    Directory of Open Access Journals (Sweden)

    Yuanyuan Nie

    2017-09-01

    Full Text Available Vessel traffic flow is a key parameter for channel-through capacity and is of great significance to vessel traffic management, channel and port design and navigational risk evaluation. Based on the study of parameters of characteristics of vessel traffic flow related to channel-through capacity, this paper puts forward a brand-new mathematical model for one-way channel-through capacity in which parameters of channel length, vessel arrival rate and velocity difference in different vessels are involved and a theoretical calculating mechanism for the channel-through capacity is provided. In order to verify availability and reliability of the model, extensive simulation studies have been carried out and based on the historical AIS data, an analytical case study on the Xiazhimen Channel validating the proposed model is presented. Both simulation studies and the case study show that the proposed model is valid and all relative parameters can be readjusted and optimized to further improve the channel-through capacity. Thus, all studies demonstrate that the model is valuable for channel design and vessel management.

  6. A revised dosimetric model of the adult head and brain

    International Nuclear Information System (INIS)

    Bouchet, L.G.; Bolch, W.E.; Weber, D.A.

    1996-01-01

    During the last decade, new radiopharmaceutical have been introduced for brain imaging. The marked differences of these tracers in tissue specificity within the brain and their increasing use for diagnostic studies support the need for a more anthropomorphic model of the human brain and head. Brain and head models developed in the past have been only simplistic representations of this anatomic region. For example, the brain within the phantom of MIRD Pamphlet No. 5 Revised is modeled simply as a single ellipsoid of tissue With no differentiation of its internal structures. To address this need, the MIRD Committee established a Task Group in 1992 to construct a more detailed brain model to include the cerebral cortex, the white matter, the cerebellum, the thalamus, the caudate nucleus, the lentiform nucleus, the cerebral spinal fluid, the lateral ventricles, and the third ventricle. This brain model has been included within a slightly modified version of the head model developed by Poston et al. in 1984. This model has been incorporated into the radiation transport code EGS4 so as to calculate photon and electron absorbed fractions in the energy range 10 keV to 4 MeV for each of thirteen sources in the brain. Furthermore, explicit positron transport have been considered, separating the contribution by the positron itself and its associated annihilations photons. No differences are found between the electron and positron absorbed fractions; however, for initial energies of positrons greater than ∼0.5 MeV, significant differences are found between absorbed fractions from explicit transport of annihilation photons and those from an assumed uniform distribution of 0.511-MeV photons. Subsequently, S values were calculated for a variety of beta-particle and positron emitters brain imaging agents. Moreover, pediatric head and brain dosimetric models are currently being developed based on this adult head model

  7. Validation of heat transfer models for gap cooling

    International Nuclear Information System (INIS)

    Okano, Yukimitsu; Nagae, Takashi; Murase, Michio

    2004-01-01

    For severe accident assessment of a light water reactor, models of heat transfer in a narrow annular gap between overheated core debris and a reactor pressure vessel are important for evaluating vessel integrity and accident management. The authors developed and improved the models of heat transfer. However, validation was not sufficient for applicability of the gap heat flux correlation to the debris cooling in the vessel lower head and applicability of the local boiling heat flux correlations to the high-pressure conditions. Therefore, in this paper, we evaluated the validity of the heat transfer models and correlations by analyses for ALPHA and LAVA experiments where molten aluminum oxide (Al 2 O 3 ) at about 2700 K was poured into the high pressure water pool in a small-scale simulated vessel lower head. In the heating process of the vessel wall, the calculated heating rate and peak temperature agreed well with the measured values, and the validity of the heat transfer models and gap heat flux correlation was confirmed. In the cooling process of the vessel wall, the calculated cooling rate was compared with the measured value, and the validity of the nucleate boiling heat flux correlation was confirmed. The peak temperatures of the vessel wall in ALPHA and LAVA experiments were lower than the temperature at the minimum heat flux point between film boiling and transition boiling, so the minimum heat flux correlation could not be validated. (author)

  8. [RESEARCH PROGRESS OF EXPERIMENTAL ANIMAL MODELS OF AVASCULAR NECROSIS OF FEMORAL HEAD].

    Science.gov (United States)

    Yu, Kaifu; Tan, Hongbo; Xu, Yongqing

    2015-12-01

    To summarize the current researches and progress on experimental animal models of avascular necrosis of the femoral head. Domestic and internation literature concerning experimental animal models of avascular necrosis of the femoral head was reviewed and analyzed. The methods to prepare the experimental animal models of avascular necrosis of the femoral head can be mainly concluded as traumatic methods (including surgical, physical, and chemical insult), and non-traumatic methods (including steroid, lipopolysaccharide, steroid combined with lipopolysaccharide, steroid combined with horse serum, etc). Each method has both merits and demerits, yet no ideal methods have been developed. There are many methods to prepare the experimental animal models of avascular necrosis of the femoral head, but proper model should be selected based on the aim of research. The establishment of ideal experimental animal models needs further research in future.

  9. 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

  10. Primo vessel inside a lymph vessel emerging from a cancer tissue.

    Science.gov (United States)

    Lee, Sungwoo; Ryu, Yeonhee; Cha, Jinmyung; Lee, Jin-Kyu; Soh, Kwang-Sup; Kim, Sungchul; Lim, Jaekwan

    2012-10-01

    Primo vessels were observed inside the lymph vessels near the caudal vena cava of a rabbit and a rat and in the thoracic lymph duct of a mouse. In the current work we found a primo vessel inside the lymph vessel that came out from the tumor tissue of a mouse. A cancer model of a nude mouse was made with human lung cancer cell line NCI-H460. We injected fluorescent nanoparticles into the xenografted tumor tissue and studied their flow in blood, lymph, and primo vessels. Fluorescent nanoparticles flowed through the blood vessels quickly in few minutes, and but slowly in the lymph vessels. The bright fluorescent signals of nanoparticles disappeared within one hour in the blood vessels but remained much longer up to several hours in the case of lymph vessels. We found an exceptional case of lymph vessels that remained bright with fluorescence up to 24 hours. After detailed examination we found that the bright fluorescence was due to a putative primo vessel inside the lymph vessel. This rare observation is consistent with Bong-Han Kim's claim on the presence of a primo vascular system in lymph vessels. It provides a significant suggestion on the cancer metastasis through primo vessels and lymph vessels. Copyright © 2012. Published by Elsevier B.V.

  11. Analysis of the procedure proposed by AREVA to prove adequate toughness of the domes of the Flamanville 3 EPR reactor pressure vessel (RPV) lower head and closure head. Session of 30 September 2015. Public version

    International Nuclear Information System (INIS)

    Catteau, R.; Cadet-Mercier, S.

    2015-01-01

    AREVA has asked ASN to evaluate the conformity of the reactor pressure vessel (RPV) for the Flamanville 3 EPR in application of the order reference [6]. The domes of the Flamanville 3 RPV closure head and lower head were manufactured in 2006 and 2007. AREVA identified that these components displayed a risk of heterogeneity of their characteristics and therefore carried out a technical qualification. At the end of 2014, AREVA informed ASN of lower-than-expected results of impact tests conducted as part of this technical qualification on test specimens taken from a dome representative of those intended for Flamanville 3. The values measured on two series of three test specimens give a mean value of 52 joules which does not attain the quality standard expected by AREVA. This mean value is also lower than the bending rupture energy value of 60 joules mentioned in point 4 of appendix 1 of the order reference [6], with which compliance would have been sufficient to prove the toughness of the material. AREVA carried out investigations to determine the origin of these noncompliant values. The carbon concentration measurements taken at the surface of the representative dome by portable spectrometry revealed the presence of a zone of major positive segregation (high concentration of carbon) over a diameter of about one meter. Furthermore, the examinations show that the segregation extends to a depth exceeding a quarter of the thickness of the dome. AREVA explains the non-compliance with the bending rupture energy criterion by the presence of this major positive segregation which came from the ingot used for the forging and was not completely eliminated by the cropping operations. To deal with this deviation, AREVA plans proving that the material is sufficiently tough by conducting new tests on a material that is representative of the lower and upper domes of the Flamanville EPR reactor. The body of the Flamanville 3 RPV, of which the lower dome is a part, has already

  12. Does Head Start differentially benefit children with risks targeted by the program's service model?

    Science.gov (United States)

    Miller, Elizabeth B; Farkas, George; Duncan, Greg J

    Data from the Head Start Impact Study ( N = 3540) were used to test for differential benefits of Head Start after one program year and after kindergarten on pre-academic and behavior outcomes for children at risk in the domains targeted by the program's comprehensive services. Although random assignment to Head Start produced positive treatment main effects on children's pre-academic skills and behavior problems, residualized growth models showed that random assignment to Head Start did not differentially benefit the pre-academic skills of children with risk factors targeted by the Head Start service model. The models showed detrimental impacts of Head Start for maternal-reported behavior problems of high-risk children, but slightly more positive impacts for teacher-reported behavior. Policy implications for Head Start are discussed.

  13. A correction on coastal heads for groundwater flow models.

    Science.gov (United States)

    Lu, Chunhui; Werner, Adrian D; Simmons, Craig T; Luo, Jian

    2015-01-01

    We introduce a simple correction to coastal heads for constant-density groundwater flow models that contain a coastal boundary, based on previous analytical solutions for interface flow. The results demonstrate that accurate discharge to the sea in confined aquifers can be obtained by direct application of Darcy's law (for constant-density flow) if the coastal heads are corrected to ((α + 1)/α)hs  - B/2α, in which hs is the mean sea level above the aquifer base, B is the aquifer thickness, and α is the density factor. For unconfined aquifers, the coastal head should be assigned the value hs1+α/α. The accuracy of using these corrections is demonstrated by consistency between constant-density Darcy's solution and variable-density flow numerical simulations. The errors introduced by adopting two previous approaches (i.e., no correction and using the equivalent fresh water head at the middle position of the aquifer to represent the hydraulic head at the coastal boundary) are evaluated. Sensitivity analysis shows that errors in discharge to the sea could be larger than 100% for typical coastal aquifer parameter ranges. The location of observation wells relative to the toe is a key factor controlling the estimation error, as it determines the relative aquifer length of constant-density flow relative to variable-density flow. The coastal head correction method introduced in this study facilitates the rapid and accurate estimation of the fresh water flux from a given hydraulic head measurement and allows for an improved representation of the coastal boundary condition in regional constant-density groundwater flow models. © 2014, National Ground Water Association.

  14. Modeling and nonlinear heading control for sailing yachts

    DEFF Research Database (Denmark)

    Xiao, Lin; Jouffroy, Jerome

    2014-01-01

    This paper presents a study on the development and testing of a model-based heading controller for a sailing yacht. Using Fossen’s compact notation for marine vehicles, we first describe a nonlinear four-degree-of-freedom (DOF) dynamic model for a sailing yacht, including roll. Our model also...

  15. Modeling and nonlinear heading control for sailing yachts

    DEFF Research Database (Denmark)

    Xiao, Lin; Jouffroy, Jerome

    2011-01-01

    This paper presents a study on the development and testing of a model-based heading controller for a sailing yacht. Using Fossen's compact notation for marine vehicles, we first describe a nonlinear 4-DOF dynamic model for a sailing yacht, including roll. Starting from this model, we then design...

  16. Instrumentation of a prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Hessheimer, M.F.; Rightley, M.J.; Matsumoto, T.

    1995-01-01

    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 U.S. Nuclear Regulatory Commission. At present, two tests are being planned: a test of a model of a steel containment vessel (SCV) that is representative of an improved, boiling water reactor (BWR) Mark II design; and a test of a model of a prestressed concrete containment vessel (PCCV). This paper discusses plans and the results of a preliminary investigation of the instrumentation of the PCCV model. The instrumentation suite for this model will consist of approximately 2000 channels of data to record displacements, strains in the reinforcing steel, prestressing tendons, concrete, steel liner and liner anchors, as well as pressure and temperature. The instrumentation is being designed to monitor the response of the model during prestressing operations, during Structural Integrity and Integrated Leak Rate testing, and during test to failure of the model. Particular emphasis has been placed on instrumentation of the prestressing system in order to understand the behavior of the prestressing strands at design and beyond design pressure levels. Current plans are to place load cells at both ends of one third of the tendons in addition to placing strain measurement devices along the length of selected tendons. Strain measurements will be made using conventional bonded foil resistance gages and a wire resistance gage, known as a open-quotes Tensmegclose quotes reg-sign gage, specifically designed for use with seven-wire strand. The results of preliminary tests of both types of gages, in the laboratory and in a simulated model configuration, are reported and plans for instrumentation of the model are discussed

  17. Nuclear reactor having an inflatable vessel closure seal structure

    International Nuclear Information System (INIS)

    1980-01-01

    An improved type of closure head seal for the rotatable plugs of the reactor vessel of a liquid metal fast breeder reactor is described. The seal prevents the release of radioactive particles while allowing the plug to be rotated without major manipulation of the seal structure. (UK)

  18. 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

  19. A Fovea Localization Scheme Using Vessel Origin-Based Parabolic Model

    Directory of Open Access Journals (Sweden)

    Chun-Yuan Yu

    2014-09-01

    Full Text Available At the center of the macula, fovea plays an important role in computer-aided diagnosis. To locate the fovea, this paper proposes a vessel origin (VO-based parabolic model, which takes the VO as the vertex of the parabola-like vasculature. Image processing steps are applied to accurately locate the fovea on retinal images. Firstly, morphological gradient and the circular Hough transform are used to find the optic disc. The structure of the vessel is then segmented with the line detector. Based on the characteristics of the VO, four features of VO are extracted, following the Bayesian classification procedure. Once the VO is identified, the VO-based parabolic model will locate the fovea. To find the fittest parabola and the symmetry axis of the retinal vessel, an Shift and Rotation (SR-Hough transform that combines the Hough transform with the shift and rotation of coordinates is presented. Two public databases of retinal images, DRIVE and STARE, are used to evaluate the proposed method. The experiment results show that the average Euclidean distances between the located fovea and the fovea marked by experts in two databases are 9.8 pixels and 30.7 pixels, respectively. The results are stronger than other methods and thus provide a better macular detection for further disease discovery.

  20. 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.

  1. Mathematical modelling for trajectories of magnetic nanoparticles in a blood vessel under magnetic field

    International Nuclear Information System (INIS)

    Sharma, Shashi; Katiyar, V.K.; Singh, Uaday

    2015-01-01

    A mathematical model is developed to describe the trajectories of a cluster of magnetic nanoparticles in a blood vessel for the application of magnetic drug targeting (MDT). The magnetic nanoparticles are injected into a blood vessel upstream from a malignant tissue and are captured at the tumour site with help of an applied magnetic field. The applied field is produced by a rare earth cylindrical magnet positioned outside the body. All forces expected to significantly affect the transport of nanoparticles were incorporated, including magnetization force, drag force and buoyancy force. The results show that particles are slow down and captured under the influence of magnetic force, which is responsible to attract the magnetic particles towards the magnet. It is optimized that all particles are captured either before or at the centre of the magnet (z≤0) when blood vessel is very close proximity to the magnet (d=2.5 cm). However, as the distance between blood vessel and magnet (d) increases (above 4.5 cm), the magnetic nanoparticles particles become free and they flow away down the blood vessel. Further, the present model results are validated by the simulations performed using the finite element based COMSOL software. - Highlights: • A mathematical model is developed to describe the trajectories of magnetic nanoparticles. • The dominant magnetic, drag and buoyancy forces are considered. • All particles are captured when distance between blood vessel and magnet (d) is up to 4.5 cm. • Further increase in d value (above 4.5 cm) results the free movement of magnetic particles

  2. 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.

  3. Patient Specific Modeling of Head-Up Tilt

    DEFF Research Database (Denmark)

    Williams, Nakeya; Wright, Andrew; Mehlsen, Jesper

    2014-01-01

    Short term cardiovascular responses to head-up tilt (HUT) experiments involve complex cardiovascular regulation in order to maintain blood pressure at homeostatic levels. This manuscript presents a patient specific compartmental model developed to predict dynamic changes in heart rate and arterial...

  4. Interpretation of strain measurements on nuclear pressure vessels

    International Nuclear Information System (INIS)

    Andersen, S.I.; Engbaek, P.

    1979-11-01

    Selected results from strain measurements on 4 nuclear pressure vessels are presented and discussed. The measurements were made in several different regions of the vessels: transition zones in vessel heads, flanges and bottom parts, nozzels, internal vessel structure and flange bolts. The results presented are based on data obtained by approximately 700 strain-gauges, and a comprehensive knowledge of the quality obtained by such measurements is established. It is shown that a thorough control procedure before and after the test as well as detailed knowledge of the behaviour of the signal from the individual gauges during the test is necessary. If this is omitted, it can be extremely difficult to distinguish between the real structural behaviour and a malfunctioning of a specific gauge installation. In general, most of the measuring results exhibit a very linear behaviour with a negligible zeroshift. However, deviations from linear behaviour are observed in several cases. This nonlinearity can be explained by friction (flange connections) or by gaps (concentrical nozzles) in certain regions, whereas local plastic deformations during the first pressure loadings of the vessel seem to be the reason in other regions. (author)

  5. A structurally detailed finite element human head model for simulation of transcranial magnetic stimulation.

    Science.gov (United States)

    Chen, Ming; Mogul, David Jeffery

    2009-04-30

    Computational studies of the head utilizing finite element models (FEMs) have been used to investigate a wide variety of brain-electromagnetic (EM) field interaction phenomena including magnetic stimulation of the head using transcranial magnetic stimulation (TMS), direct electric stimulation of the brain for electroconvulsive therapy, and electroencephalography source localization. However, no human head model of sufficient complexity for studying the biophysics under these circumstances has been developed which utilizes structures at both the regional and cellular levels and provides well-defined smooth boundaries between tissues of different conductivities and orientations. The main barrier for building such accurate head models is the complex modeling procedures that include 3D object reconstruction and optimized meshing. In this study, a structurally detailed finite element model of the human head was generated that includes details to the level of cerebral gyri and sulci by combining computed tomography and magnetic resonance images. Furthermore, cortical columns that contain conductive processes of pyramidal neurons traversing the neocortical layers were included in the head model thus providing structure at or near the cellular level. These refinements provide a much more realistic model to investigate the effects of TMS on brain electrophysiology in the neocortex.

  6. Final report for the 'Melt-Vessel Interactions' Project. European Union R and TD Program 4th Framework. MVI project final research report

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Dinh, T.N.; Nourgaliev, R.R.; Bui, V.A.; Green, J.; Kolb, G.; Karbojian, A.; Theerthan, S.A.; Gubaidulline, A.; Bonnet, J.M.; Rouge, S.; Narcoux, M.; Liegeois, A.; Turland, B.D.; Dobson, G.P.; Siccama, A.; Ikonen, K.; Parozzi, F.; Kolev, N.; Caira, M.

    1999-04-01

    The Melt Vessel Interaction (MVI) project is concerned with the consequences of the interactions that a core melt, generated during a postulated severe accident in a light water reactor, may have with the pressure vessel. In particular, the issues concerned with the failure of the vessel bottom head are the focus of the research. The specific objectives of the project are to obtain data and develop validated models, which could be applied to prototypic plants, and accident conditions, for resolution of issues related to the melt vessel interactions. The project work has been performed by nine partners having varied responsibility. The work included a large number of experiments, with simulant materials, whose observations and results are employed, respectively, to understand the physical mechanisms and to develop validated models. Applications to the prototypic geometry and conditions have also been performed. This report is volume 1 of the Final Report for the Project, in which a summary of the progress achieved in the experimental program is provided. We have, however, included some aspects of the modeling activities. Volume 2 of the Final report describes the progress achieved in the modeling program. The progress achieved in the experimental and modeling parts of the Project has led to the resolution of some of the issues of melt vessel interaction. Considerable progress was also achieved towards resolution of the remaining issues

  7. Estimation of heading gyrocompass error using a GPS 3DF system: Impact on ADCP measurements

    Directory of Open Access Journals (Sweden)

    Simón Ruiz

    2002-12-01

    Full Text Available Traditionally the horizontal orientation in a ship (heading has been obtained from a gyrocompass. This instrument is still used on research vessels but has an estimated error of about 2-3 degrees, inducing a systematic error in the cross-track velocity measured by an Acoustic Doppler Current Profiler (ADCP. The three-dimensional positioning system (GPS 3DF provides an independent heading measurement with accuracy better than 0.1 degree. The Spanish research vessel BIO Hespérides has been operating with this new system since 1996. For the first time on this vessel, the data from this new instrument are used to estimate gyrocompass error. The methodology we use follows the scheme developed by Griffiths (1994, which compares data from the gyrocompass and the GPS system in order to obtain an interpolated error function. In the present work we apply this methodology on mesoscale surveys performed during the observational phase of the OMEGA project, in the Alboran Sea. The heading-dependent gyrocompass error dominated. Errors in gyrocompass heading of 1.4-3.4 degrees have been found, which give a maximum error in measured cross-track ADCP velocity of 24 cm s-1.

  8. Animation of 3D Model of Human Head

    Directory of Open Access Journals (Sweden)

    V. Michalcin

    2007-04-01

    Full Text Available The paper deals with the new algorithm of animation of 3D model of the human head in combination with its global motion. The designed algorithm is very fast and with low calculation requirements, because it does not need the synthesis of the input videosequence for estimation of the animation parameters as well as the parameters of global motion. The used 3D model Candide generates different expressions using its animation units which are controlled by the animation parameters. These ones are estimated on the basis of optical flow without the need of extracting of the feature points in the frames of the input videosequence because they are given by the selected vertices of the animation units of the calibrated 3D model Candide. The established multiple iterations inside the designed animation algorithm of 3D model of the human head between two successive frames significantly improved its accuracy above all for the large motion.

  9. Anti-lymphangiogenic properties of mTOR inhibitors in head and neck squamous cell carcinoma experimental models

    International Nuclear Information System (INIS)

    Ekshyyan, Oleksandr; Moore-Medlin, Tara N; Raley, Matthew C; Sonavane, Kunal; Rong, Xiaohua; Brodt, Michael A; Abreo, Fleurette; Alexander, Jonathan Steven; Nathan, Cherie-Ann O

    2013-01-01

    Tumor dissemination to cervical lymph nodes via lymphatics represents the first step in the metastasis of head and neck squamous cell carcinoma (HNSCC) and is the most significant predictor of tumor recurrence decreasing survival by 50%. The lymphatic suppressing properties of mTOR inhibitors are not yet well understood. Lymphatic inhibiting effects of rapamycin were evaluated in vitro using two lymphatic endothelial cell (LEC) lines. An orthotopic mouse model of HNSCC (OSC-19 cells) was used to evaluate anti-lymphangiogenic effects of rapamycin in vivo. The incidence of cervical lymph node metastases, numbers of tumor-free lymphatic vessels and those invaded by tumor cells in mouse lingual tissue, and expression of pro-lymphangiogenic markers were assessed. Rapamycin significantly decreased lymphatic vascular density (p = 0.027), reduced the fraction of lymphatic vessels invaded by tumor cells in tongue tissue (p = 0.013) and decreased metastasis-positive lymph nodes (p = 0.04). Rapamycin also significantly attenuated the extent of metastatic tumor cell spread within lymph nodes (p < 0.0001). We found that rapamycin significantly reduced LEC proliferation and was correlated with decreased VEGFR-3 expression in both LEC, and in some HNSCC cell lines. The results of this study demonstrate anti-lymphangiogenic properties of mTOR inhibitors in HNSCC. mTOR inhibitors suppress autocrine and paracrine growth stimulation of tumor and lymphatic endothelial cells by impairing VEGF-C/VEGFR-3 axis and release of soluble VEGFR-2. In a murine HNSCC orthotopic model rapamycin significantly suppressed lymphovascular invasion, decreased cervical lymph node metastasis and delayed the spread of metastatic tumor cells within the lymph nodes

  10. Applied model of through-wall crack of coolant vessels of WWER-type reactors

    International Nuclear Information System (INIS)

    Petrosyan, V.; Hovakimyan, T.; Vardanyan, M.; Khachatryan, A.; Minasyan, K.

    2010-01-01

    We propose an applied-model of Through-Wall Crack (TWC) for WWER-type units primary vessels. The model allows to simulate the main morphological parameters of real TWC, i.e. length, area of inlet and outlet openings, channel depth and small and large size unevenness of the crack surface. The model can be used for developing and improving the coolant-leak detectors for the primary circuit vessels of WWER-units. Also, it can be used for research of the coolant two-phase leakage phenomenon through narrow cracks/channels and thermo-physical processes in heat-insulation layer of the Main Coolant Piping (MCP) during the leak

  11. An agent-based model of the response to angioplasty and bare-metal stent deployment in an atherosclerotic blood vessel.

    Directory of Open Access Journals (Sweden)

    Antonia E Curtin

    Full Text Available PURPOSE: While animal models are widely used to investigate the development of restenosis in blood vessels following an intervention, computational models offer another means for investigating this phenomenon. A computational model of the response of a treated vessel would allow investigators to assess the effects of altering certain vessel- and stent-related variables. The authors aimed to develop a novel computational model of restenosis development following an angioplasty and bare-metal stent implantation in an atherosclerotic vessel using agent-based modeling techniques. The presented model is intended to demonstrate the body's response to the intervention and to explore how different vessel geometries or stent arrangements may affect restenosis development. METHODS: The model was created on a two-dimensional grid space. It utilizes the post-procedural vessel lumen diameter and stent information as its input parameters. The simulation starting point of the model is an atherosclerotic vessel after an angioplasty and stent implantation procedure. The model subsequently generates the final lumen diameter, percent change in lumen cross-sectional area, time to lumen diameter stabilization, and local concentrations of inflammatory cytokines upon simulation completion. Simulation results were directly compared with the results from serial imaging studies and cytokine levels studies in atherosclerotic patients from the relevant literature. RESULTS: The final lumen diameter results were all within one standard deviation of the mean lumen diameters reported in the comparison studies. The overlapping-stent simulations yielded results that matched published trends. The cytokine levels remained within the range of physiological levels throughout the simulations. CONCLUSION: We developed a novel computational model that successfully simulated the development of restenosis in a blood vessel following an angioplasty and bare-metal stent deployment based on

  12. On path generation and feedforward control for a class of surface sailing vessels

    DEFF Research Database (Denmark)

    Xiao, Lin; Jouffroy, Jerome

    2010-01-01

    Sailing vessels with wind as their main means of propulsion possess a unique property that the paths they take depend on the wind direction, which, in the literature, has attracted less attention than normal vehicles propelled by propellers or thrusters. This paper considers the problem of motion...... planning and controllability for sailing vehicles representing the no-sailing zone effect in sailing. Following our previous work, we present an extended algorithm for automatic path generation with a prescribed initial heading for a simple model of sailing vehicles, together with a feedforward controller...

  13. Significance of head CT in neuro-ophthalmology

    International Nuclear Information System (INIS)

    Nishimoto, Yuichiro; Masuyama, Yoshimasa; Nakamura, Yasuko; Fujimoto, Toshiro.

    1979-01-01

    It is important to perform CT purposefully in patients with some neuro-ophthalmological disorders. Using Hitachi CT-H250 with the finer matrix of 256 x 256 and a section thickness of 5 mm and 10 mm, we performed head CT in 98 patients with various kinds of neuro-ophthalmological disorders in these 30 months. Neuro-ophthalmological findings in these patients were visual disturbance, visual field defects (hemianopsia, quandrantanopsia, enlargement of the blind spot, central scotoma), papilledema, optic nerve atrophy, difficulties in ocular movements, and others. In 44 (44.9%) of these 98 patients, abnormal findings were displayed in the head CT, characterizing intracranial tumors, intracranial infarction, aneurysms in intracranial vessels, arterio-venous malformation, enlargement of the ventricle or the cistern. In some cases the head CT was the decisive procedure in ensuring a correct diagnosis. We presented findings in the head CT in several interesting cases. We recognized the necessity of the head CT in patients with neuro-ophthalmological disorders. (author)

  14. Hydraulic head interpolation using ANFIS—model selection and sensitivity analysis

    Science.gov (United States)

    Kurtulus, Bedri; Flipo, Nicolas

    2012-01-01

    The aim of this study is to investigate the efficiency of ANFIS (adaptive neuro fuzzy inference system) for interpolating hydraulic head in a 40-km 2 agricultural watershed of the Seine basin (France). Inputs of ANFIS are Cartesian coordinates and the elevation of the ground. Hydraulic head was measured at 73 locations during a snapshot campaign on September 2009, which characterizes low-water-flow regime in the aquifer unit. The dataset was then split into three subsets using a square-based selection method: a calibration one (55%), a training one (27%), and a test one (18%). First, a method is proposed to select the best ANFIS model, which corresponds to a sensitivity analysis of ANFIS to the type and number of membership functions (MF). Triangular, Gaussian, general bell, and spline-based MF are used with 2, 3, 4, and 5 MF per input node. Performance criteria on the test subset are used to select the 5 best ANFIS models among 16. Then each is used to interpolate the hydraulic head distribution on a (50×50)-m grid, which is compared to the soil elevation. The cells where the hydraulic head is higher than the soil elevation are counted as "error cells." The ANFIS model that exhibits the less "error cells" is selected as the best ANFIS model. The best model selection reveals that ANFIS models are very sensitive to the type and number of MF. Finally, a sensibility analysis of the best ANFIS model with four triangular MF is performed on the interpolation grid, which shows that ANFIS remains stable to error propagation with a higher sensitivity to soil elevation.

  15. Drill core investigations from the TMI-2 pressure vessel. Final report

    International Nuclear Information System (INIS)

    Sturm, D.; Katerbau, K.H.; Maile, K.; Ruoff, H.

    1994-01-01

    For the evaluation of the results obtained in TMI-2 VIP and for the preparation of the continuing discussion in the OECD and of research measures in the national sphere but also for the appraisal of the effect of the results to date on safety philosophy and safety research in Germany, the present research project, inter alia, was commenced. In content was: a) Furtherance of the OECD-NEA-TMI-2 Vessel Investigation Project in dealing with the testing programme by active collaboration in the Programme Review Group, by participation in ad-hoc meetings on the question of specimen extraction, by advice on the conduct of metallographic, metallurgical and mechanical investigations on the specimens from the RPV bottom head and by assessment of the findings. b) Investigation of specimens from the bottom head of the TMI-2 reactor pressure vessel. c) Investigation of specimens from archive material. The investigations reach the widely agreed conclusion that during the accident a hot spot developed in the bottom head of the reactor in which for a time of about 30 minutes a maximum temperature of some 1100 C or greater than 900 C prevailed. Around this zone there is a region with temperatures higher than ca. 730 C (A 1 ) whilst the predominant portion of the head had not been heated beyond the 1 temperature. (orig.) [de

  16. Prediction of thermoplastic failure of a reactor pressure vessel under a postulated core melt accident

    International Nuclear Information System (INIS)

    Duijvestijn, G.; Birchley, J.; Reichlin, K.

    1997-01-01

    This paper presents the lower head failure calculations performed for a postulated accident scenario in a commercial nuclear power plant. A postulated one inch break in the primary coolant circuit leads to dryout and subsequent meltdown of the core. The reference plant is a pressurized water reactor without penetrations in the reactor vessel lower head. The molten core material accumulates in the lower head, eventually causing failure of the vessel. The analysis investigates flow conditions in the melt pool, temperature evolution in the reactor vessel wall, and structure mechanical evaluation of the vessel under strong thermal loads and a range of internal pressures. The calculations were performed using the ADINA finite element codes. The analysis focusses on the failure processes, time and mode of failure. The most likely mode of failure at low pressure is global rupture due to gradual accumulation of creep strain over a large part of the heated area. In contrast, thermoplasticity becomes important at high pressure or following a pressure spike and can lead to earlier local failure. In situations in which part of the heat load is concentrated over a small area, resulting in a hot spot, local failure occurs, but not until the temperatures are close to the melting point. At low pressure, in particular, the hot spot area remains intact until the structure is molten across more than half of the thickness. (author) 14 figs., 16 refs

  17. Lower Length Scale Model Development for Embrittlement of Reactor Presure Vessel Steel

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report summarizes the lower-length-scale effort during FY 2016 in developing mesoscale capabilities for microstructure evolution, plasticity and fracture in reactor pressure vessel steels. During operation, reactor pressure vessels are subject to hardening and embrittlement caused by irradiation induced defect accumulation and irradiation enhanced solute precipitation. Both defect production and solute precipitation start from the atomic scale, and manifest their eventual effects as degradation in engineering scale properties. To predict the property degradation, multiscale modeling and simulation are needed to deal with the microstructure evolution, and to link the microstructure feature to material properties. In this report, the development of mesoscale capabilities for defect accumulation and solute precipitation are summarized. A crystal plasticity model to capture defect-dislocation interaction and a damage model for cleavage micro-crack propagation is also provided.

  18. Correcting electrode modelling errors in EIT on realistic 3D head models.

    Science.gov (United States)

    Jehl, Markus; Avery, James; Malone, Emma; Holder, David; Betcke, Timo

    2015-12-01

    Electrical impedance tomography (EIT) is a promising medical imaging technique which could aid differentiation of haemorrhagic from ischaemic stroke in an ambulance. One challenge in EIT is the ill-posed nature of the image reconstruction, i.e., that small measurement or modelling errors can result in large image artefacts. It is therefore important that reconstruction algorithms are improved with regard to stability to modelling errors. We identify that wrongly modelled electrode positions constitute one of the biggest sources of image artefacts in head EIT. Therefore, the use of the Fréchet derivative on the electrode boundaries in a realistic three-dimensional head model is investigated, in order to reconstruct electrode movements simultaneously to conductivity changes. We show a fast implementation and analyse the performance of electrode position reconstructions in time-difference and absolute imaging for simulated and experimental voltages. Reconstructing the electrode positions and conductivities simultaneously increased the image quality significantly in the presence of electrode movement.

  19. Glaucoma Diagnostic Ability of the Optical Coherence Tomography Angiography Vessel Density Parameters.

    Science.gov (United States)

    Chung, Jae Keun; Hwang, Young Hoon; Wi, Jae Min; Kim, Mijin; Jung, Jong Jin

    2017-11-01

    To investigate the glaucoma diagnostic abilities of vessel density parameters as determined by optical coherence tomography (OCT) angiography in different stages of glaucoma. A total of 113 healthy eyes and 140 glaucomatous eyes were enrolled. Diagnostic abilities of the OCT vessel density parameters in the optic nerve head (ONH), peripapillary, and macular regions were evaluated by calculating the area under the receiver operation characteristic curves (AUCs). AUCs of the peripapillary vessel density parameters and circumpapillary retinal nerve fiber layer (RNFL) thickness were compared. OCT angiography vessel densities in the ONH, peripapillary, and macular regions in the glaucomatous eyes were significantly lower than those in the healthy eyes (P glaucoma detection. The peripapillary vessel density parameters showed similar AUCs with the corresponding sectoral RNFL thickness (P > 0.05). However, in the early stage of glaucoma, the AUCs of the inferotemporal and temporal peripapillary vessel densities were significantly lower than that of the RNFL thickness (P glaucoma diagnostic ability with circumpapillary RNFL thickness, in the early stage, the vessel density parameters showed limited clinical value.

  20. Nonlinear response of vessel walls due to short-time thermomechanical loading

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kulak, R.F.

    1994-01-01

    Maintaining structural integrity of the reactor pressure vessel (RPV) during a postulated core melt accident is an important safety consideration in the design of the vessel. This study addresses the failure predictions of the vessel due to thermal and pressure loadings fro the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on the dead load, yield stress assumptions, material response and internal pressurization. The analyses considered only short term failure (quasi static) modes, long term failure modes were not considered. Short term failure modes include plastic instabilities of the structure and failure due to exceeding the failure strain. Long term failure odes would be caused by creep rupture that leads to plastic instability of the structure. Due to the sort time durations analyzed, creep was not considered in the analyses presented

  1. Prestressed concrete reactor vessel thermal cylinder model study

    International Nuclear Information System (INIS)

    Callahan, J.P.; Canonico, D.A.; Richardson, M.; Corum, J.M.; Dodge, W.G.; Robinson, G.C.; Whitman, G.D.

    1977-01-01

    The thermal cylinder experiment was designed both to provide information for evaluating the capability of analytical methods to predict the time-dependent stress-strain behavior of a 1 / 6 -scale model of the barrel section of a single-cavity prestressed concrete reactor vessel and to demonstrate the structural behavior under design and off-design thermal conditions. The model was a thick-walled cylinder having a height of 1.22 m, a thickness of 0.46 m, and an outer diameter of 2.06 m. It was prestressed both axially and circumferentially and subjected to 4.83 MPa internal pressure together with a thermal crossfall imposed by heating the inner surface to 338.8 K and cooling the outer surface to 297.1 K. The initial 460 days of testing were divided into time periods that simulated prestressing, heatup, reactor operation, and shutdown. At the conclusion of the simulated operating period, the model was repressurized and subjected to localized heating at 505.4 K for 84 days to produce an off-design hot-spot condition. Comparisons of experimental data with calculated values obtained using the SAFE-CRACK finite-element computer program showed that the program was capable of predicting time-dependent behavior in a vessel subjected to normal operating conditions, but that it was unable to accurately predict the behavior during off-design hot-spot heating. Readings made using a neutron and gamma-ray backscattering moisture probe showed little, if any, migration of moisture in the concrete cross section. Destructive examination indicated that the model maintained its basic structural integrity during localized hot-spot heating

  2. 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

  3. Combining operational models and data into a dynamic vessel risk assessment tool for coastal regions

    Science.gov (United States)

    Fernandes, R.; Braunschweig, F.; Lourenço, F.; Neves, R.

    2016-02-01

    The technological evolution in terms of computational capacity, data acquisition systems, numerical modelling and operational oceanography is supplying opportunities for designing and building holistic approaches and complex tools for newer and more efficient management (planning, prevention and response) of coastal water pollution risk events. A combined methodology to dynamically estimate time and space variable individual vessel accident risk levels and shoreline contamination risk from ships has been developed, integrating numerical metocean forecasts and oil spill simulations with vessel tracking automatic identification systems (AIS). The risk rating combines the likelihood of an oil spill occurring from a vessel navigating in a study area - the Portuguese continental shelf - with the assessed consequences to the shoreline. The spill likelihood is based on dynamic marine weather conditions and statistical information from previous accidents. The shoreline consequences reflect the virtual spilled oil amount reaching shoreline and its environmental and socio-economic vulnerabilities. The oil reaching shoreline is quantified with an oil spill fate and behaviour model running multiple virtual spills from vessels along time, or as an alternative, a correction factor based on vessel distance from coast. Shoreline risks can be computed in real time or from previously obtained data. Results show the ability of the proposed methodology to estimate the risk properly sensitive to dynamic metocean conditions and to oil transport behaviour. The integration of meteo-oceanic + oil spill models with coastal vulnerability and AIS data in the quantification of risk enhances the maritime situational awareness and the decision support model, providing a more realistic approach in the assessment of shoreline impacts. The risk assessment from historical data can help finding typical risk patterns ("hot spots") or developing sensitivity analysis to specific conditions, whereas real

  4. A concern about the crack propagation rate of PWSCC which obtained from the investigation on primary coolant leakage portion of the reactor vessel head in Ohi 3

    International Nuclear Information System (INIS)

    Totsuka, Nobuo; Fukumura, Takuya

    2010-01-01

    There will be some concern about the content presented in the paper entitled 'Primary Coolant Leakage Path Research of Reactor Vessel Head Penetration' published in INSS JOURNAL of 2008, which may lead to misunderstanding about the PWSCC crack propagation rate, that is, the rate written in the paper seems to be faster than those reported by the previous studies. It is considered that such misunderstanding will be due to a sentence in the abstract of the paper. Therefore, we will revise a part of the abstract and explain about the outline of the paper again. (author)

  5. Design of Hemispherical Downward-Facing Vessel for Critical Heat Flux Experiment

    International Nuclear Information System (INIS)

    Hwang, J. S.; Suh, K. Y.

    2009-01-01

    The in-vessel retention (IVR) is one of major severe accident management strategies adopted by some operating nuclear power plants during a severe accident. The recent Shin-Gori Units 3 and 4 of the Advanced Power Reactor 1400 MWe (APR1400) have adopted the external reactor vessel cooling (ERVC) by reactor cavity flooding as major severe accident management strategy. The ERVC in the APR1400 design resorts to active flooding system using thermal insulator. The Corium Attack Stopper Apparatus Spherical Channel (CASA SC) tests are conducted to measure the critical power and critical heat flux (CHF) on a downward hemispherical vessel scaled down from the APR1400 lower head by 1/10 on a linear scale. CASA is designed through scaling and thermal analysis to simulate the APR1400 vessel and thermal insulator. The heated vessel of CASA SC represents the external surface of a hemisphere submerged vessel in water. The heated vessel plays an important role in the ERVC experiment depending on the configuration of oxide pool and metallic layer. Hand calculation and computational analysis are performed to produce high heat flux from the downward facing hemisphere in excess of 1 MW/m 2

  6. Final report for the 'Melt-Vessel Interactions' Project. European Union R and TD Program 4th Framework. MVI project final research report

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Dinh, T.N.; Nourgaliev, R.R.; Bui, V.A.; Green, J.; Kolb, G.; Karbojian, A.; Theerthan, S.A.; Gubaidulline, A. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety; Helle, M.; Kymaelaeinen, O.; Tuomisto, H. [IVO Power Engineering Ltd., Vantaa (Finland); Bonnet, J.M.; Rouge, S.; Narcoux, M.; Liegeois, A. [CEA - Grenoble (France); Turland, B.D.; Dobson, G.P. [AEA Technology plc, Dorchester (United Kingdom); Siccama, A. [ECN Nuclear Research, Petten (Netherlands); Ikonen, K. [VTT Energy, Helsinki (Finland); Parozzi, F. [ENEL - SRI/PAM/GRA, Segrate, MI (Italy); Kolev, N. [Siemens AG, Erlangen (Germany); Caira, M. [Univ. of Roma (Italy)

    1999-04-01

    The Melt Vessel Interaction (MVI) project is concerned with the consequences of the interactions that a core melt, generated during a postulated severe accident in a light water reactor, may have with the pressure vessel. In particular, the issues concerned with the failure of the vessel bottom head are the focus of the research. The specific objectives of the project are to obtain data and develop validated models, which could be applied to prototypic plants, and accident conditions, for resolution of issues related to the melt vessel interactions. The project work has been performed by nine partners having varied responsibility. The work included a large number of experiments, with simulant materials, whose observations and results are employed, respectively, to understand the physical mechanisms and to develop validated models. Applications to the prototypic geometry and conditions have also been performed. This report is volume 1 of the Final Report for the Project, in which a summary of the progress achieved in the experimental program is provided. We have, however, included some aspects of the modeling activities. Volume 2 of the Final report describes the progress achieved in the modeling program. The progress achieved in the experimental and modeling parts of the Project has led to the resolution of some of the issues of melt vessel interaction. Considerable progress was also achieved towards resolution of the remaining issues.

  7. Comparison of a networks-of-zones fluid mixing model for a baffled stirred vessel with three-dimensional electrical resistance tomography

    International Nuclear Information System (INIS)

    Rodgers, T L; Siperstein, F R; Mann, R; York, T A; Kowalski, A

    2011-01-01

    Reliable models for the simulation of mixing vessels are important for the understanding of real-life mixing problems. To achieve these models, information about the mixing in the system must be measured to compare with the predicted values. Electrical resistance tomography has the capability to measure spatial and temporal changes within a vessel in three dimensions even in optically inaccessible environments. This paper discusses the creation of a network-of-zones model for the prediction of mixing within a vessel with a Cowles disc-type agitator. Solving of the network-of-zones simplified transport equations for the vessel predicts the concentration distribution of an inert tracer added to the vessel. The change in this distribution with time is calculated and compared with visual inspection of the vessel. The concentration distribution inside the vessel is also measured using electrical resistance tomography and shows good agreement with the predicted values

  8. Rapidly re-computable EEG (electroencephalography) forward models for realistic head shapes

    International Nuclear Information System (INIS)

    Ermer, J.J.; Mosher, J.C.; Baillet, S.; Leahy, R.M.

    2001-01-01

    Solution of the EEG source localization (inverse) problem utilizing model-based methods typically requires a significant number of forward model evaluations. For subspace based inverse methods like MUSIC (6), the total number of forward model evaluations can often approach an order of 10 3 or 10 4 . Techniques based on least-squares minimization may require significantly more evaluations. The observed set of measurements over an M-sensor array is often expressed as a linear forward spatio-temporal model of the form: F = GQ + N (1) where the observed forward field F (M-sensors x N-time samples) can be expressed in terms of the forward model G, a set of dipole moment(s) Q (3xP-dipoles x N-time samples) and additive noise N. Because of their simplicity, ease of computation, and relatively good accuracy, multi-layer spherical models (7) (or fast approximations described in (1), (7)) have traditionally been the 'forward model of choice' for approximating the human head. However, approximation of the human head via a spherical model does have several key drawbacks. By its very shape, the use of a spherical model distorts the true distribution of passive currents in the skull cavity. Spherical models also require that the sensor positions be projected onto the fitted sphere (Fig. 1), resulting in a distortion of the true sensor-dipole spatial geometry (and ultimately the computed surface potential). The use of a single 'best-fitted' sphere has the added drawback of incomplete coverage of the inner skull region, often ignoring areas such as the frontal cortex. In practice, this problem is typically countered by fitting additional sphere(s) to those region(s) not covered by the primary sphere. The use of these additional spheres results in added complication to the forward model. Using high-resolution spatial information obtained via X-ray CT or MR imaging, a realistic head model can be formed by tessellating the head into a set of contiguous regions (typically the scalp

  9. Habitat Suitability Index Models: Yellow-headed blackbird

    Science.gov (United States)

    Schroeder, Richard L.

    1982-01-01

    Habitat preferences of the yellow-headed blackbird (Xanthocephalus xanthocephalus) are described in this publication. It is one of a series of Habitat Suitability Index (HSI) models and was developed through an analysis of available infomration on the species-habitat requirements of the species. Habitat use information is presented in a review of the literature, followed by the development of an HSI model, designed for use in impact assessment and habitat management activities.

  10. 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.

  11. Walking Ahead: The Headed Social Force Model.

    Directory of Open Access Journals (Sweden)

    Francesco Farina

    Full Text Available Human motion models are finding an increasing number of novel applications in many different fields, such as building design, computer graphics and robot motion planning. The Social Force Model is one of the most popular alternatives to describe the motion of pedestrians. By resorting to a physical analogy, individuals are assimilated to point-wise particles subject to social forces which drive their dynamics. Such a model implicitly assumes that humans move isotropically. On the contrary, empirical evidence shows that people do have a preferred direction of motion, walking forward most of the time. Lateral motions are observed only in specific circumstances, such as when navigating in overcrowded environments or avoiding unexpected obstacles. In this paper, the Headed Social Force Model is introduced in order to improve the realism of the trajectories generated by the classical Social Force Model. The key feature of the proposed approach is the inclusion of the pedestrians' heading into the dynamic model used to describe the motion of each individual. The force and torque representing the model inputs are computed as suitable functions of the force terms resulting from the traditional Social Force Model. Moreover, a new force contribution is introduced in order to model the behavior of people walking together as a single group. The proposed model features high versatility, being able to reproduce both the unicycle-like trajectories typical of people moving in open spaces and the point-wise motion patterns occurring in high density scenarios. Extensive numerical simulations show an increased regularity of the resulting trajectories and confirm a general improvement of the model realism.

  12. Macroscopic and microscopic findings in avascular necrosis of the femoral head.

    Science.gov (United States)

    Kamal, Diana; Alexandru, D O; Kamal, C K; Streba, C T; Grecu, D; Mogoantă, L

    2012-01-01

    The avascular necrosis of the femoral head is an illness induced by the cutoff of blood flow to the femoral head and it affects mostly young adults between the ages of 30 and 50 years, raising therapeutic and diagnostic issues. Many risk factors are incriminated in the development of avascular necrosis of the femoral head like: trauma, chronic alcohol consumption, smoking, administration of corticosteroid drugs, most of the cases are considered to be idiopathic. The main goal of our paper is to describe the macroscopic and microscopic variations of the bone structure, which occur in patients with avascular necrosis of the femoral head. The biological material needed for our study was obtained following hip arthroplasty surgery in 26 patients between the ages of 29 and 59 years, which previously were diagnosed with avascular necrosis of the femoral head and admitted in the Orthopedics Department of the Emergency County Hospital of Craiova (Romania) between 2010 and 2011. From a macroscopic point of view, we found well defined areas of necrosis, most of which were neatly demarcated of the adjacent viable tissue by hyperemic areas, loss of shape and contour of the femoral head and transformations of the articular cartilage above the area of necrosis. When examined under the microscope, we found vast areas of fibrosis, narrow bone trabeculae, obstructed blood vessels or blood vessels with clots inside, hypertrophic fat cells, bone sequestration but also small cells and pyknotic nuclei. The microscopic and macroscopic findings on the femoral head sections varied with the patients and the stage of the disease.

  13. An Approach for Selection of Flow Regime and Models for Conservative Evaluation of a Vessel Integrity Monitoring System for Water-Cooled Vacuum Vessels

    International Nuclear Information System (INIS)

    Pointer, W. David; Ruggles, Arthur E.

    2003-01-01

    Thin-walled vacuum containment vessels cooled by circulating water jackets are often utilized in research and industrial applications where isolation of equipment or experiments from the influences of the surrounding environment is desirable. The development of leaks in these vessels can result in costly downtime for the facility. A Vessel Integrity Monitoring System (VIMS) is developed to detect leak formation and estimate the size of the leak to allow evaluation of the risk associated with continued operation. A wide range of leak configurations and fluid flow phenomena are considered in the evaluation of the rate at which a tracer gas dissolved in the cooling jacket water is transported into the vacuum vessel. A methodology is presented that uses basic fluid flow models and careful evaluation of their ranges of applicability to provide a conservative estimate of the transport rates for the tracer gas and hence the time required for the VIMS to detect a leak of a given size

  14. Development of TPNCIRC code for Evaluation of Two-Phase Natural Circulation Flow Performance under External Reactor Vessel Cooling Conditions

    International Nuclear Information System (INIS)

    Choi, A-Reum; Song, Hyuk-Jin; Park, Jong-Woon

    2015-01-01

    During a severe accident, corium is relocated to the lower head of the nuclear reactor pressure vessel (RPV). Design concept of retaining the corium inside a nuclear reactor pressure vessel (RPV) through external cooling under hypothetical core melting accidents is called external reactor vessel cooling (ERVC). In this respect, validated two-phase natural circulation flow (TPNC) model is necessary to determine the adequacy of the ERVC design and operating conditions such as inlet area, form losses, gap distance, riser length and coolant conditions. The most important model generally characterizing the TPNC are void fraction and two-phase friction factors. Typical experimental and analytical studies to be referred to on two-phase circulation flow characteristics are those by Reyes, Gartia et al. based on Vijayan et al., Nayak et al. and Dubey et al. In the present paper, two-phase natural circulation (TPNC) flow characteristics under external reactor vessel cooling (ERVC) conditions are studied using two existing TPNC flow models of Reyes and Gartia et al. incorporating more improved void fraction and two-phase friction models. These models and correlations are integrated into a computer program, TPNCIRC, which can handle candidate ERVC design parameters, such as inlet, riser and downcomer flow lengths and areas, gap size between reactor vessel and surrounding insulations, minor loss factors and operating parameters of decay power, pressure and subcooling. Accuracy of the TPNCIRC program is investigated with respect to the flow rate and void fractions for existing measured data from a general experiment and ULPU specifically designed for the AP1000 in-vessel retention. Also, the effect of some important design parameters are examined for the experimental and plant conditions. Using the flow models and correlations are integrated into a computer program, TPNCIRC, a number of correlations have been examined. This seems coming from the differences of void fractions

  15. 1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Kim, Jae Cheol

    2007-02-01

    During a severe accident, when a molten corium is relocated in a reactor vessel lower head, the RCF(Reactor Cavity Flooding) system for ERVC (External Reactor Vessel Cooling) is actuated and coolants are supplied into a reactor cavity to remove a decay heat from the molten corium. This severe accident mitigation strategy for maintaining a integrity of reactor vessel was adopted in the nuclear power plants of APR1400, AP600, and AP1000. Under the ERVC condition, the upward two-phase flow is driven by the amount of the decay heat from the molten corium. To achieve the ERVC strategy, the two-phase natural circulation in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. Also the natural circulation flow restriction has to be minimized. In this reason, it is needed to review the fundamental structure of insulation. In the existing power plants, the insulation design is aimed at minimizing heat losses under a normal operation. Under the ERVC condition, however, the ability to form the two-phase natural circulation is uncertain. Namely, some important factors, such as the coolant inlet/outlet areas, flow restriction, and steam vent etc. in the flow channel, should be considered for ERVC design. T-HEMES 1D study is launched to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The air injection method was used to simulate the boiling at the external reactor vessel and generate the natural circulation two-phase flow. From the experimental results, the natural circulation flow rate highly depended on inlet/outlet areas and the circulation flow rate increased as the outlet height as well as the supplied water head increased. On the other hand, the simple analysis using the drift

  16. Does Head Start differentially benefit children with risks targeted by the program’s service model?☆

    Science.gov (United States)

    Miller, Elizabeth B.; Farkas, George; Duncan, Greg J.

    2015-01-01

    Data from the Head Start Impact Study (N = 3540) were used to test for differential benefits of Head Start after one program year and after kindergarten on pre-academic and behavior outcomes for children at risk in the domains targeted by the program’s comprehensive services. Although random assignment to Head Start produced positive treatment main effects on children’s pre-academic skills and behavior problems, residualized growth models showed that random assignment to Head Start did not differentially benefit the pre-academic skills of children with risk factors targeted by the Head Start service model. The models showed detrimental impacts of Head Start for maternal-reported behavior problems of high-risk children, but slightly more positive impacts for teacher-reported behavior. Policy implications for Head Start are discussed. PMID:26379369

  17. Computational Fluid Dynamics Analysis of Pulsatile Blood Flow Behavior in Modelled Stenosed Vessels with Different Severities

    Directory of Open Access Journals (Sweden)

    Mohsen Mehrabi

    2012-01-01

    Full Text Available This study focuses on the behavior of blood flow in the stenosed vessels. Blood is modelled as an incompressible non-Newtonian fluid which is based on the power law viscosity model. A numerical technique based on the finite difference method is developed to simulate the blood flow taking into account the transient periodic behaviour of the blood flow in cardiac cycles. Also, pulsatile blood flow in the stenosed vessel is based on the Womersley model, and fluid flow in the lumen region is governed by the continuity equation and the Navier-Stokes equations. In this study, the stenosis shape is cosine by using Tu and Devil model. Comparing the results obtained from three stenosed vessels with 30%, 50%, and 75% area severity, we find that higher percent-area severity of stenosis leads to higher extrapressure jumps and higher blood speeds around the stenosis site. Also, we observe that the size of the stenosis in stenosed vessels does influence the blood flow. A little change on the cross-sectional value makes vast change on the blood flow rate. This simulation helps the people working in the field of physiological fluid dynamics as well as the medical practitioners.

  18. Skull Defects in Finite Element Head Models for Source Reconstruction from Magnetoencephalography Signals

    Science.gov (United States)

    Lau, Stephan; Güllmar, Daniel; Flemming, Lars; Grayden, David B.; Cook, Mark J.; Wolters, Carsten H.; Haueisen, Jens

    2016-01-01

    Magnetoencephalography (MEG) signals are influenced by skull defects. However, there is a lack of evidence of this influence during source reconstruction. Our objectives are to characterize errors in source reconstruction from MEG signals due to ignoring skull defects and to assess the ability of an exact finite element head model to eliminate such errors. A detailed finite element model of the head of a rabbit used in a physical experiment was constructed from magnetic resonance and co-registered computer tomography imaging that differentiated nine tissue types. Sources of the MEG measurements above intact skull and above skull defects respectively were reconstructed using a finite element model with the intact skull and one incorporating the skull defects. The forward simulation of the MEG signals reproduced the experimentally observed characteristic magnitude and topography changes due to skull defects. Sources reconstructed from measured MEG signals above intact skull matched the known physical locations and orientations. Ignoring skull defects in the head model during reconstruction displaced sources under a skull defect away from that defect. Sources next to a defect were reoriented. When skull defects, with their physical conductivity, were incorporated in the head model, the location and orientation errors were mostly eliminated. The conductivity of the skull defect material non-uniformly modulated the influence on MEG signals. We propose concrete guidelines for taking into account conducting skull defects during MEG coil placement and modeling. Exact finite element head models can improve localization of brain function, specifically after surgery. PMID:27092044

  19. Adaptable three-dimensional Monte Carlo modeling of imaged blood vessels in skin

    Science.gov (United States)

    Pfefer, T. Joshua; Barton, Jennifer K.; Chan, Eric K.; Ducros, Mathieu G.; Sorg, Brian S.; Milner, Thomas E.; Nelson, J. Stuart; Welch, Ashley J.

    1997-06-01

    In order to reach a higher level of accuracy in simulation of port wine stain treatment, we propose to discard the typical layered geometry and cylindrical blood vessel assumptions made in optical models and use imaging techniques to define actual tissue geometry. Two main additions to the typical 3D, weighted photon, variable step size Monte Carlo routine were necessary to achieve this goal. First, optical low coherence reflectometry (OLCR) images of rat skin were used to specify a 3D material array, with each entry assigned a label to represent the type of tissue in that particular voxel. Second, the Monte Carlo algorithm was altered so that when a photon crosses into a new voxel, the remaining path length is recalculated using the new optical properties, as specified by the material array. The model has shown good agreement with data from the literature. Monte Carlo simulations using OLCR images of asymmetrically curved blood vessels show various effects such as shading, scattering-induced peaks at vessel surfaces, and directionality-induced gradients in energy deposition. In conclusion, this augmentation of the Monte Carlo method can accurately simulate light transport for a wide variety of nonhomogeneous tissue geometries.

  20. Movement of retinal vessels toward the optic nerve head after increasing intraocular pressure in monkey eyes with experimental glaucoma.

    Science.gov (United States)

    Kuroda, Atsumi; Enomoto, Nobuko; Ishida, Kyoko; Shimazawa, Masamitsu; Noguchi, Tetsuro; Horai, Naoto; Onoe, Hirotaka; Hara, Hideaki; Tomita, Goji

    2017-09-01

    A shift or displacement of the retinal blood vessels (RBVs) with neuroretinal rim thinning indicates the progression of glaucomatous optic neuropathy. In chronic open angle glaucoma, individuals with RBV positional shifts exhibit more rapid visual field loss than those without RBV shifts. The retinal vessels reportedly move onto the optic nerve head (ONH) in response to glaucoma damage, suggesting that RBVs are pulled toward the ONH in response to increased cupping. Whether this phenomenon only applies to RVBs located in the vicinity or inside the ONH or, more generally, to RBVs also located far from the ONH, however, is unclear. The aim of this study was to evaluate the movement of RBVs located relatively far from the ONH edge after increasing intraocular pressure (IOP) in an experimental monkey model of glaucoma. Fundus photographs were obtained in 17 monkeys. High IOP was induced in the monkeys by laser photocoagulation burns applied uniformly with 360° irradiation around the trabecular meshwork of the left eye. The right eye was left intact and used as a non-treated control. Considering the circadian rhythm of IOP, it was measured in both eyes of each animal at around the same time-points. Then, fundus photographs were obtained. Using Image J image analysis software, an examiner (N.E.) measured the fundus photographs at two time-points, i.e. before laser treatment (time 1) and the last fundus photography after IOP elevation (time 2). The following parameters were measured (in pixels): 1) vertical diameter of the ONH (DD), 2) distance from the ONH edge to the first bifurcation point of the superior branch of the central retinal vein (UV), 3) distance from the ONH edge to the first bifurcation point of the inferior branch of the central retinal vein (LV), 4) ONH area, and 5) surface area of the cup of the ONH. We calculated the ratios of UV to DD (UV/DD), LV to DD (LV/DD), and the cup area to disc area ratio (C/D). The mean UV/DD at time 1 (0.656 ± 0.233) was

  1. Corrected Four-Sphere Head Model for EEG Signals.

    Science.gov (United States)

    Næss, Solveig; Chintaluri, Chaitanya; Ness, Torbjørn V; Dale, Anders M; Einevoll, Gaute T; Wójcik, Daniel K

    2017-01-01

    The EEG signal is generated by electrical brain cell activity, often described in terms of current dipoles. By applying EEG forward models we can compute the contribution from such dipoles to the electrical potential recorded by EEG electrodes. Forward models are key both for generating understanding and intuition about the neural origin of EEG signals as well as inverse modeling, i.e., the estimation of the underlying dipole sources from recorded EEG signals. Different models of varying complexity and biological detail are used in the field. One such analytical model is the four-sphere model which assumes a four-layered spherical head where the layers represent brain tissue, cerebrospinal fluid (CSF), skull, and scalp, respectively. While conceptually clear, the mathematical expression for the electric potentials in the four-sphere model is cumbersome, and we observed that the formulas presented in the literature contain errors. Here, we derive and present the correct analytical formulas with a detailed derivation. A useful application of the analytical four-sphere model is that it can serve as ground truth to test the accuracy of numerical schemes such as the Finite Element Method (FEM). We performed FEM simulations of the four-sphere head model and showed that they were consistent with the corrected analytical formulas. For future reference we provide scripts for computing EEG potentials with the four-sphere model, both by means of the correct analytical formulas and numerical FEM simulations.

  2. Corrected Four-Sphere Head Model for EEG Signals

    Directory of Open Access Journals (Sweden)

    Solveig Næss

    2017-10-01

    Full Text Available The EEG signal is generated by electrical brain cell activity, often described in terms of current dipoles. By applying EEG forward models we can compute the contribution from such dipoles to the electrical potential recorded by EEG electrodes. Forward models are key both for generating understanding and intuition about the neural origin of EEG signals as well as inverse modeling, i.e., the estimation of the underlying dipole sources from recorded EEG signals. Different models of varying complexity and biological detail are used in the field. One such analytical model is the four-sphere model which assumes a four-layered spherical head where the layers represent brain tissue, cerebrospinal fluid (CSF, skull, and scalp, respectively. While conceptually clear, the mathematical expression for the electric potentials in the four-sphere model is cumbersome, and we observed that the formulas presented in the literature contain errors. Here, we derive and present the correct analytical formulas with a detailed derivation. A useful application of the analytical four-sphere model is that it can serve as ground truth to test the accuracy of numerical schemes such as the Finite Element Method (FEM. We performed FEM simulations of the four-sphere head model and showed that they were consistent with the corrected analytical formulas. For future reference we provide scripts for computing EEG potentials with the four-sphere model, both by means of the correct analytical formulas and numerical FEM simulations.

  3. Testing the dual-route model of perceived gaze direction: Linear combination of eye and head cues.

    Science.gov (United States)

    Otsuka, Yumiko; Mareschal, Isabelle; Clifford, Colin W G

    2016-06-01

    We have recently proposed a dual-route model of the effect of head orientation on perceived gaze direction (Otsuka, Mareschal, Calder, & Clifford, 2014; Otsuka, Mareschal, & Clifford, 2015), which computes perceived gaze direction as a linear combination of eye orientation and head orientation. By parametrically manipulating eye orientation and head orientation, we tested the adequacy of a linear model to account for the effect of horizontal head orientation on perceived direction of gaze. Here, participants adjusted an on-screen pointer toward the perceived gaze direction in two image conditions: Normal condition and Wollaston condition. Images in the Normal condition included a change in the visible part of the eye along with the change in head orientation, while images in the Wollaston condition were manipulated to have identical eye regions across head orientations. Multiple regression analysis with explanatory variables of eye orientation and head orientation revealed that linear models account for most of the variance both in the Normal condition and in the Wollaston condition. Further, we found no evidence that the model with a nonlinear term explains significantly more variance. Thus, the current study supports the dual-route model that computes the perceived gaze direction as a linear combination of eye orientation and head orientation.

  4. Experimental test of spatial updating models for monkey eye-head gaze shifts.

    Directory of Open Access Journals (Sweden)

    Tom J Van Grootel

    Full Text Available How the brain maintains an accurate and stable representation of visual target locations despite the occurrence of saccadic gaze shifts is a classical problem in oculomotor research. Here we test and dissociate the predictions of different conceptual models for head-unrestrained gaze-localization behavior of macaque monkeys. We adopted the double-step paradigm with rapid eye-head gaze shifts to measure localization accuracy in response to flashed visual stimuli in darkness. We presented the second target flash either before (static, or during (dynamic the first gaze displacement. In the dynamic case the brief visual flash induced a small retinal streak of up to about 20 deg at an unpredictable moment and retinal location during the eye-head gaze shift, which provides serious challenges for the gaze-control system. However, for both stimulus conditions, monkeys localized the flashed targets with accurate gaze shifts, which rules out several models of visuomotor control. First, these findings exclude the possibility that gaze-shift programming relies on retinal inputs only. Instead, they support the notion that accurate eye-head motor feedback updates the gaze-saccade coordinates. Second, in dynamic trials the visuomotor system cannot rely on the coordinates of the planned first eye-head saccade either, which rules out remapping on the basis of a predictive corollary gaze-displacement signal. Finally, because gaze-related head movements were also goal-directed, requiring continuous access to eye-in-head position, we propose that our results best support a dynamic feedback scheme for spatial updating in which visuomotor control incorporates accurate signals about instantaneous eye- and head positions rather than relative eye- and head displacements.

  5. Accelerated Life Structural Benchmark Testing for a Stirling Convertor Heater Head

    Science.gov (United States)

    Krause, David L.; Kantzos, Pete T.

    2006-01-01

    For proposed long-duration NASA Space Science missions, the Department of Energy, Lockheed Martin, Infinia Corporation, and NASA Glenn Research Center are developing a high-efficiency, 110 W Stirling Radioisotope Generator (SRG110). A structurally significant limit state for the SRG110 heater head component is creep deformation induced at high material temperature and low stress level. Conventional investigations of creep behavior adequately rely on experimental results from uniaxial creep specimens, and a wealth of creep data is available for the Inconel 718 material of construction. However, the specified atypical thin heater head material is fine-grained with a heat treatment that limits precipitate growth, and little creep property data for this microstructure is available in the literature. In addition, the geometry and loading conditions apply a multiaxial stress state on the component, far from the conditions of uniaxial testing. For these reasons, an extensive experimental investigation is ongoing to aid in accurately assessing the durability of the SRG110 heater head. This investigation supplements uniaxial creep testing with pneumatic testing of heater head-like pressure vessels at design temperature with stress levels ranging from approximately the design stress to several times that. This paper presents experimental results, post-test microstructural analyses, and conclusions for four higher-stress, accelerated life tests. Analysts are using these results to calibrate deterministic and probabilistic analytical creep models of the SRG110 heater head.

  6. Polynomial fuzzy model-based approach for underactuated surface vessels

    DEFF Research Database (Denmark)

    Khooban, Mohammad Hassan; Vafamand, Navid; Dragicevic, Tomislav

    2018-01-01

    The main goal of this study is to introduce a new polynomial fuzzy model-based structure for a class of marine systems with non-linear and polynomial dynamics. The suggested technique relies on a polynomial Takagi–Sugeno (T–S) fuzzy modelling, a polynomial dynamic parallel distributed compensation...... surface vessel (USV). Additionally, in order to overcome the USV control challenges, including the USV un-modelled dynamics, complex nonlinear dynamics, external disturbances and parameter uncertainties, the polynomial fuzzy model representation is adopted. Moreover, the USV-based control structure...... and a sum-of-squares (SOS) decomposition. The new proposed approach is a generalisation of the standard T–S fuzzy models and linear matrix inequality which indicated its effectiveness in decreasing the tracking time and increasing the efficiency of the robust tracking control problem for an underactuated...

  7. A Huge Capital Drop with Compression of Femoral Vessels Associated with Hip Osteoarthritis

    Directory of Open Access Journals (Sweden)

    Tomoya Takasago

    2015-01-01

    Full Text Available A capital drop is a type of osteophyte at the inferomedial portion of the femoral head commonly observed in hip osteoarthritis (OA, secondary to developmental dysplasia. Capital drop itself is typically asymptomatic; however, symptoms can appear secondary to impinge against the acetabulum or to irritation of the surrounding tissues, such as nerves, vessels, and tendons. We present here a case of unilateral leg edema in a patient with hip OA, caused by a huge bone mass occurring at the inferomedial portion of the femoral head that compressed the femoral vessels. We diagnosed this bone mass as a capital drop secondary to hip OA after confirming that the mass occurred at least after the age of 63 years based on a previous X-ray. We performed early resection and total hip arthroplasty since the patient’s hip pain was due to both advanced hip OA and compression of the femoral vessels; moreover, we aimed to prevent venous thrombosis secondary to vascular compression considering the advanced age and the potent risk of thrombosis in the patient. A large capital drop should be considered as a cause of vascular compression in cases of unilateral leg edema in OA patients.

  8. 46 CFR 32.55-5 - Ventilation of tank vessels constructed between November 10, 1936, and July 1, 1951-TB/ALL.

    Science.gov (United States)

    2010-10-01

    ... actuated gas ejectors or blowers or ventilators fitted with heads for natural ventilation, will be approved... 46 Shipping 1 2010-10-01 2010-10-01 false Ventilation of tank vessels constructed between November... HOMELAND SECURITY TANK VESSELS SPECIAL EQUIPMENT, MACHINERY, AND HULL REQUIREMENTS Ventilation and Venting...

  9. Realistic Avatar Eye and Head Animation Using a Neurobiological Model of Visual Attention

    National Research Council Canada - National Science Library

    Itti, L; Dhavale, N; Pighin, F

    2003-01-01

    We describe a neurobiological model of visual attention and eye/head movements in primates, and its application to the automatic animation of a realistic virtual human head watching an unconstrained...

  10. Development, Validation and Parametric study of a 3-Year-Old Child Head Finite Element Model

    Science.gov (United States)

    Cui, Shihai; Chen, Yue; Li, Haiyan; Ruan, ShiJie

    2015-12-01

    Traumatic brain injury caused by drop and traffic accidents is an important reason for children's death and disability. Recently, the computer finite element (FE) head model has been developed to investigate brain injury mechanism and biomechanical responses. Based on CT data of a healthy 3-year-old child head, the FE head model with detailed anatomical structure was developed. The deep brain structures such as white matter, gray matter, cerebral ventricle, hippocampus, were firstly created in this FE model. The FE model was validated by comparing the simulation results with that of cadaver experiments based on reconstructing the child and adult cadaver experiments. In addition, the effects of skull stiffness on the child head dynamic responses were further investigated. All the simulation results confirmed the good biofidelity of the FE model.

  11. A stochastic-bayesian model for the fracture probability of PWR pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Francisco, Alexandre S.; Duran, Jorge Alberto R., E-mail: afrancisco@metal.eeimvr.uff.br, E-mail: duran@metal.eeimvr.uff.br [Universidade Federal Fluminense (UFF), Volta Redonda, RJ (Brazil). Dept. de Engenharia Mecanica

    2013-07-01

    Fracture probability of pressure vessels containing cracks can be obtained by methodologies of easy understanding, which require a deterministic treatment, complemented by statistical methods. However, more accurate results are required, methodologies need to be better formulated. This paper presents a new methodology to address this problem. First, a more rigorous methodology is obtained by means of the relationship of probability distributions that model crack incidence and nondestructive inspection efficiency using the Bayes' theorem. The result is an updated crack incidence distribution. Further, the accuracy of the methodology is improved by using a stochastic model for the crack growth. The stochastic model incorporates the statistical variability of the crack growth process, combining the stochastic theory with experimental data. Stochastic differential equations are derived by the randomization of empirical equations. From the solution of this equation, a distribution function related to the crack growth is derived. The fracture probability using both probability distribution functions is in agreement with theory, and presents realistic value for pressure vessels. (author)

  12. A stochastic-bayesian model for the fracture probability of PWR pressure vessels

    International Nuclear Information System (INIS)

    Francisco, Alexandre S.; Duran, Jorge Alberto R.

    2013-01-01

    Fracture probability of pressure vessels containing cracks can be obtained by methodologies of easy understanding, which require a deterministic treatment, complemented by statistical methods. However, more accurate results are required, methodologies need to be better formulated. This paper presents a new methodology to address this problem. First, a more rigorous methodology is obtained by means of the relationship of probability distributions that model crack incidence and nondestructive inspection efficiency using the Bayes' theorem. The result is an updated crack incidence distribution. Further, the accuracy of the methodology is improved by using a stochastic model for the crack growth. The stochastic model incorporates the statistical variability of the crack growth process, combining the stochastic theory with experimental data. Stochastic differential equations are derived by the randomization of empirical equations. From the solution of this equation, a distribution function related to the crack growth is derived. The fracture probability using both probability distribution functions is in agreement with theory, and presents realistic value for pressure vessels. (author)

  13. Experiments on performance of the multi-layered in-vessel core catcher

    International Nuclear Information System (INIS)

    Kang, K.H.; Kim, S.B.; Park, R.J.; Cheung, F.B.; Suh, K.Y.; Rempe, J.L.

    2004-01-01

    LAVA-GAP experiments are in progress to investigate the performance of the in-vessel core catcher using alumina melt as a corium simulant. The hemispherical in-vessel core catcher made of carbon steel was installed inside the lower head vessel with a uniform gap of 10 mm. Until now, two types of the in-vessel core catcher were used in this study. The first one is a single layered in-vessel core catcher without an internal coating of the LAVA-GAP-2 test, and the other one is a two layered in-vessel core catcher with a 0.5 mm-thick ZrO 2 internal coating of the LAVA-GAP-3 test. Current LAVA-GAP experimental results indicate that an internally coated in-vessel core catcher has better thermal performance compared with an uncoated in-vessel core catcher. Metallurgical inspections on the test specimens of the LAVA-GAP-3 test have been performed to examine the performance of the coating material and the base carbon steel. Although the base carbon steel had experienced a severe thermal attack to the extent that the microstructures were changed and re-crystallization occurred, the carbon steel showed stable and pure chemical compositions without any oxidation and interaction with the coating layer. In terms of the material aspects, these metallurgical inspection results suggest that the ZrO 2 coating performed well. (authors)

  14. Development of Realistic Head Models for Electromagnetic Source Imaging of the Human Brain

    National Research Council Canada - National Science Library

    Akalin, Z

    2001-01-01

    In this work, a methodology is developed to solve the forward problem of electromagnetic source imaging using realistic head models, For this purpose, first segmentation of the 3 dimensional MR head...

  15. Comparison of a layered slab and an atlas head model for Monte Carlo fitting of time-domain near-infrared spectroscopy data of the adult head.

    Science.gov (United States)

    Selb, Juliette; Ogden, Tyler M; Dubb, Jay; Fang, Qianqian; Boas, David A

    2014-01-01

    Near-infrared spectroscopy (NIRS) estimations of the adult brain baseline optical properties based on a homogeneous model of the head are known to introduce significant contamination from extracerebral layers. More complex models have been proposed and occasionally applied to in vivo data, but their performances have never been characterized on realistic head structures. Here we implement a flexible fitting routine of time-domain NIRS data using graphics processing unit based Monte Carlo simulations. We compare the results for two different geometries: a two-layer slab with variable thickness of the first layer and a template atlas head registered to the subject's head surface. We characterize the performance of the Monte Carlo approaches for fitting the optical properties from simulated time-resolved data of the adult head. We show that both geometries provide better results than the commonly used homogeneous model, and we quantify the improvement in terms of accuracy, linearity, and cross-talk from extracerebral layers.

  16. Pictorial essay: Vascular interventions in extra cranial head and neck

    Directory of Open Access Journals (Sweden)

    Suyash S Kulkarni

    2012-01-01

    Full Text Available Medicine is an ever changing field and interventional radiology (IR procedures are becoming increasingly popular because of high efficacy and its minimally invasive nature of the procedure. Management of disease processes in the extra cranial head and neck (ECHN has always been a challenge due to the complex anatomy of the region. Cross sectional imaging of the ECHN has grown and evolved tremendously and occupies a pivotal and integral position in the clinical management of variety of head and neck pathologies. Advances in angiographic technologies including flat panel detector systems, biplane, and 3-dimensional rotational angiography have consolidated and expanded the role of IR in the management of various ECHN pathologies. The ECHN is at cross roads between the origins of great vessels and the cerebral vasculature. Thorough knowledge of functional and technical aspects of neuroangiography is essential before embarking on head and neck vascular interventions. The vessels of the head and neck can be involved by infectious and inflammatory conditions, get irradiated during radiotherapy and injured due to trauma or iatrogenic cause. The ECHN is also a common site for various hypervascular neoplasms and vascular malformations, which can be treated with endovascular and percutaneous embolization. This pictorial essay provides a review of variety of ECHN pathologies which were managed by various IR procedures using different approaches.

  17. Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

    Energy Technology Data Exchange (ETDEWEB)

    Houry, M., E-mail: Michael.houry@cea.fr [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France); Gargiulo, L.; Balorin, C.; Bruno, V.; Keller, D.; Roche, H. [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France); Kammerer, N.; Measson, Y. [CEA, LIST, F-92265 Fontenay-aux-Roses (France); Carrel, F.; Schoepff, V. [CEA, LIST, F-91191 Gif-sur-Yvette (France)

    2011-10-15

    ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.

  18. Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

    International Nuclear Information System (INIS)

    Houry, M.; Gargiulo, L.; Balorin, C.; Bruno, V.; Keller, D.; Roche, H.; Kammerer, N.; Measson, Y.; Carrel, F.; Schoepff, V.

    2011-01-01

    ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.

  19. 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

  20. 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

  1. Markov Chain Model-Based Optimal Cluster Heads Selection for Wireless Sensor Networks

    Directory of Open Access Journals (Sweden)

    Gulnaz Ahmed

    2017-02-01

    Full Text Available The longer network lifetime of Wireless Sensor Networks (WSNs is a goal which is directly related to energy consumption. This energy consumption issue becomes more challenging when the energy load is not properly distributed in the sensing area. The hierarchal clustering architecture is the best choice for these kind of issues. In this paper, we introduce a novel clustering protocol called Markov chain model-based optimal cluster heads (MOCHs selection for WSNs. In our proposed model, we introduce a simple strategy for the optimal number of cluster heads selection to overcome the problem of uneven energy distribution in the network. The attractiveness of our model is that the BS controls the number of cluster heads while the cluster heads control the cluster members in each cluster in such a restricted manner that a uniform and even load is ensured in each cluster. We perform an extensive range of simulation using five quality measures, namely: the lifetime of the network, stable and unstable region in the lifetime of the network, throughput of the network, the number of cluster heads in the network, and the transmission time of the network to analyze the proposed model. We compare MOCHs against Sleep-awake Energy Efficient Distributed (SEED clustering, Artificial Bee Colony (ABC, Zone Based Routing (ZBR, and Centralized Energy Efficient Clustering (CEEC using the above-discussed quality metrics and found that the lifetime of the proposed model is almost 1095, 2630, 3599, and 2045 rounds (time steps greater than SEED, ABC, ZBR, and CEEC, respectively. The obtained results demonstrate that the MOCHs is better than SEED, ABC, ZBR, and CEEC in terms of energy efficiency and the network throughput.

  2. A mathematical model for batch and continuous thickening of flocculent suspensions in vessels with varying section

    Energy Technology Data Exchange (ETDEWEB)

    Buerger, R.; Damasceno, J.J.R.; Karlesen, K.H.

    2001-10-01

    The phenomenological theory of continuous thickening of flocculated suspensions in an ideal cylindrical thickener is extended to vessels having varying cross-section, including divergent or convergent conical vessels. The purpose of this contribution is to draw attention to the corresponding mathematical model, whose key ingredient is a strongly degenerate parabolic partial differential equation. For ideal (non-flocculated) suspensions, which do not form co compressible sediments, the mathematical model reduces to the kinematic approach by Anestis, who developed a method of construction of exact solution by the method of characteristics. The difficulty lies in the fact that characteristics and iso-concentration lines, unlike the conventional Kynch model for cylindrical vessels, do not coincide, and one has to resort to numerical methods to simulate the thickening process. A numerical algorithm is presented and employed for simulations of continuous thickening. Implications of the mathematical model are also demonstrated by steady-state calculations, which lead to new possibilities in thickener design. (author)

  3. Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

    Energy Technology Data Exchange (ETDEWEB)

    Wang, C.Y.

    1993-06-01

    This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts` ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

  4. Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

    Energy Technology Data Exchange (ETDEWEB)

    Wang, C.Y.

    1993-01-01

    This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts' ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

  5. Vascular patterns in the heads of crocodilians: blood vessels and sites of thermal exchange.

    Science.gov (United States)

    Porter, William Ruger; Sedlmayr, Jayc C; Witmer, Lawrence M

    2016-12-01

    Extant crocodilians are a highly apomorphic archosaur clade that is ectothermic, yet often achieve large body sizes that can be subject to higher heat loads. Therefore, the anatomical and physiological roles that blood vessels play in crocodilian thermoregulation need further investigation to better understand how crocodilians establish and maintain cephalic temperatures and regulate neurosensory tissue temperatures during basking and normal activities. The cephalic vascular anatomy of extant crocodilians, particularly American alligator (Alligator mississippiensis) was investigated using a differential-contrast, dual-vascular injection technique and high resolution X-ray micro-computed tomography (μCT). Blood vessels were digitally isolated to create representations of vascular pathways. The specimens were then dissected to confirm CT results. Sites of thermal exchange, consisting of the oral, nasal, and orbital regions, were given special attention due to their role in evaporative cooling and cephalic thermoregulation in other diapsids. Blood vessels to and from sites of thermal exchange were studied to detect conserved vascular patterns and to assess their ability to deliver cooled blood to neurosensory tissues. Within the orbital region, both the arteries and veins demonstrated consistent branching patterns, with the supraorbital, infraorbital, and ophthalmotemporal vessels supplying and draining the orbit. The venous drainage of the orbital region showed connections to the dural sinuses via the orbital veins and cavernous sinus. The palatal region demonstrated a vast plexus that comprised both arteries and veins. The most direct route of venous drainage of the palatal plexus was through the palatomaxillary veins, essentially bypassing neurosensory tissues. Anastomotic connections with the nasal region, however, may provide an alternative route for palatal venous blood to reach neurosensory tissues. The nasal region in crocodilians is probably the most

  6. Quench of molten aluminum oxide associated with in-vessel debris retention by RPV internal water

    International Nuclear Information System (INIS)

    Maruyama, Yu; Yamano, Norihiro; Moriyama, Kiyofumi; Park, Hyun Sun; Kudo, Tamotsu; Yang, Yanhua; Sugimoto, Jun

    1999-01-01

    In-vessel debris coolability experiments were performed in ALPHA program at JAERI. Molten aluminum oxide (Al 2 O 3 ) was poured into a pool of water in a lower head experimental vessel. Post-test observation and measurement using an ultrasonic technique indicated the formation of the interfacial gap between the solidified Al 2 O 3 and the vessel wall. Thermal responses of the vessel wall implied that the interfacial gap acted initially as a thermal resistance and water subsequently penetrated into the interfacial gap. The maximum heat flux at the inner surface of the vessel facing to the solidified Al 2 O 3 was roughly evaluated to be ranged from 320 kW/m 2 to 600 kW/m 2 . A post-test analysis was conducted with CAMP code. The influence of the interfacial gap on thermal behavior of Al 2 O 3 and the vessel wall was examined. (authors)

  7. Modeling of melt retention in EU-APR1400 ex-vessel core catcher

    Energy Technology Data Exchange (ETDEWEB)

    Granovsky, V. S.; Sulatsky, A. A.; Khabensky, V. B.; Sulatskaya, M. B. [Alexandrov Research Inst. of Technology NITI, Sosnovy Bor (Russian Federation); Gusarov, V. V.; Almyashev, V. I.; Komlev, A. A. [Saint Petersburg State Technological Univ. SPbSTU, St.Petersburg (Russian Federation); Bechta, S. [KTH, Stockholm (Sweden); Kim, Y. S. [KHNP, 1312 Gil 70, Yuseongdaero, Yuseong-gu, Daejeon (Korea, Republic of); Park, R. J.; Kim, H. Y.; Song, J. H. [KAERI, 989 Gil 111, Daedeokdaero, Yuseong-gu, Daejeon (Korea, Republic of)

    2012-07-01

    A core catcher is adopted in the EU-APR1400 reactor design for management and mitigation of severe accidents with reactor core melting. The core catcher concept incorporates a number of engineering solutions used in the catcher designs of European EPR and Russian WER-1000 reactors, such as thin-layer corium spreading for better cooling, retention of the melt in a water-cooled steel vessel, and use of sacrificial material (SM) to control the melt properties. SM is one of the key elements of the catcher design and its performance is critical for melt retention efficiency. This SM consists of oxide components, but the core catcher also includes sacrificial steel which reacts with the metal melt of the molten corium to reduce its temperature. The paper describes the required properties of SM. The melt retention capability of the core catcher can be confirmed by modeling the heat fluxes to the catcher vessel to show that it will not fail. The fulfillment of this requirement is demonstrated on the example of LBLOCA severe accident. Thermal and physicochemical interactions between the oxide and metal melts, interactions of the melts with SM, sacrificial steel and vessel, core catcher external cooling by water and release of non-condensable gases are modeled. (authors)

  8. Vessel guardians: sculpture and graphics related to the ceramics of NorthEastern European hunter-gatherers

    Directory of Open Access Journals (Sweden)

    Ekaterina Aleksandrovna Kashina

    2015-12-01

    Full Text Available North-Eastern European hunter-gatherer ceramic sculptures, relief sculptures and graphic images on vessels are discussed. Five groups of finds are distinguished according to their chronology (4000–2500 BC cal and represented subject (birds, human head, human figure, mammal head etc.. Their production believes to be a female craft, their making had ritual aims and their emerging was independent from any influences of pastoral/agricultural societies.

  9. Modelling of hydrogen deflagration in a vented vessel

    International Nuclear Information System (INIS)

    Wang, L.L.; Wong, R.C.

    1995-01-01

    Hydrogen Deflagration inside closed and vented 2.3 m diameter vessels were simulated by using the GOTHIC lumped-parameter computer code. Different cell arrangements were used in the modelling. Other parameters such as flame speed and hydrogen concentration were studied. It was found that the calculated peak pressures for cases using the experimental measured burn durations were close to the pressures measured from the experiments. When the default flame speed was used, higher peak pressure was predicted by GOTHIC. This could be explained by the the fact that the default flame speed used in the GOTHIC burn model was based on the results of a large scale test with moderate turbulence level. However, the overall results of the pressure transients were comparable with the experimental data. In addition, time and spatial convergencies of the model were also studied. The peak pressure estimated by modelling the sphere as five or more spherical cells was shown to converge to within +/- 3 percent. (author). 8 refs., 6 tabs., 9 figs

  10. [Virtual audiovisual talking heads: articulatory data and models--applications].

    Science.gov (United States)

    Badin, P; Elisei, F; Bailly, G; Savariaux, C; Serrurier, A; Tarabalka, Y

    2007-01-01

    In the framework of experimental phonetics, our approach to the study of speech production is based on the measurement, the analysis and the modeling of orofacial articulators such as the jaw, the face and the lips, the tongue or the velum. Therefore, we present in this article experimental techniques that allow characterising the shape and movement of speech articulators (static and dynamic MRI, computed tomodensitometry, electromagnetic articulography, video recording). We then describe the linear models of the various organs that we can elaborate from speaker-specific articulatory data. We show that these models, that exhibit a good geometrical resolution, can be controlled from articulatory data with a good temporal resolution and can thus permit the reconstruction of high quality animation of the articulators. These models, that we have integrated in a virtual talking head, can produce augmented audiovisual speech. In this framework, we have assessed the natural tongue reading capabilities of human subjects by means of audiovisual perception tests. We conclude by suggesting a number of other applications of talking heads.

  11. Principle Study of Head Meridian Acupoint Massage to Stress Release via Grey Data Model Analysis.

    Science.gov (United States)

    Lee, Ya-Ting

    2016-01-01

    This paper presents the scientific study of the effectiveness and action principle of head meridian acupoint massage by applying the grey data model analysis approach. First, the head massage procedure for massaging the important head meridian acupuncture points including Taiyang, Fengfu, Tianzhu, Fengqi, and Jianjing is formulated in a standard manner. Second, the status of the autonomic nervous system of each subject is evaluated by using the heart rate variability analyzer before and after the head massage following four weeks. Afterward, the physiological factors of autonomic nerves are quantitatively analyzed by using the grey data modeling theory. The grey data analysis can point out that the status of autonomic nervous system is greatly improved after the massage. The order change of the grey relationship weighting of physiological factors shows the action principle of the sympathetic and parasympathetic nerves when performing head massage. In other words, the grey data model is able to distinguish the detailed interaction of the autonomic nervous system and the head meridian acupoint massage. Thus, the stress relaxing effect of massaging head meridian acupoints is proved, which is lacked in literature. The results can be a reference principle for massage health care in practice.

  12. 3D realistic head model simulation based on transcranial magnetic stimulation.

    Science.gov (United States)

    Yang, Shuo; Xu, Guizhi; Wang, Lei; Chen, Yong; Wu, Huanli; Li, Ying; Yang, Qingxin

    2006-01-01

    Transcranial magnetic stimulation (TMS) is a powerful non-invasive tool for investigating functions in the brain. The target inside the head is stimulated with eddy currents induced in the tissue by the time-varying magnetic field. Precise spatial localization of stimulation sites is the key of efficient functional magnetic stimulations. Many researchers devote to magnetic field analysis in empty free space. In this paper, a realistic head model used in Finite Element Method has been developed. The magnetic field inducted in the head bt TMS has been analysed. This three-dimensional simulation is useful for spatial localization of stimulation.

  13. Comparison of analytic source models for head scatter factor calculation and planar dose calculation for IMRT

    International Nuclear Information System (INIS)

    Yan Guanghua; Liu, Chihray; Lu Bo; Palta, Jatinder R; Li, Jonathan G

    2008-01-01

    The purpose of this study was to choose an appropriate head scatter source model for the fast and accurate independent planar dose calculation for intensity-modulated radiation therapy (IMRT) with MLC. The performance of three different head scatter source models regarding their ability to model head scatter and facilitate planar dose calculation was evaluated. A three-source model, a two-source model and a single-source model were compared in this study. In the planar dose calculation algorithm, in-air fluence distribution was derived from each of the head scatter source models while considering the combination of Jaw and MLC opening. Fluence perturbations due to tongue-and-groove effect, rounded leaf end and leaf transmission were taken into account explicitly. The dose distribution was calculated by convolving the in-air fluence distribution with an experimentally determined pencil-beam kernel. The results were compared with measurements using a diode array and passing rates with 2%/2 mm and 3%/3 mm criteria were reported. It was found that the two-source model achieved the best agreement on head scatter factor calculation. The three-source model and single-source model underestimated head scatter factors for certain symmetric rectangular fields and asymmetric fields, but similar good agreement could be achieved when monitor back scatter effect was incorporated explicitly. All the three source models resulted in comparable average passing rates (>97%) when the 3%/3 mm criterion was selected. The calculation with the single-source model and two-source model was slightly faster than the three-source model due to their simplicity

  14. Comparison of analytic source models for head scatter factor calculation and planar dose calculation for IMRT

    Energy Technology Data Exchange (ETDEWEB)

    Yan Guanghua [Department of Nuclear and Radiological Engineering, University of Florida, Gainesville, FL 32611 (United States); Liu, Chihray; Lu Bo; Palta, Jatinder R; Li, Jonathan G [Department of Radiation Oncology, University of Florida, Gainesville, FL 32610-0385 (United States)

    2008-04-21

    The purpose of this study was to choose an appropriate head scatter source model for the fast and accurate independent planar dose calculation for intensity-modulated radiation therapy (IMRT) with MLC. The performance of three different head scatter source models regarding their ability to model head scatter and facilitate planar dose calculation was evaluated. A three-source model, a two-source model and a single-source model were compared in this study. In the planar dose calculation algorithm, in-air fluence distribution was derived from each of the head scatter source models while considering the combination of Jaw and MLC opening. Fluence perturbations due to tongue-and-groove effect, rounded leaf end and leaf transmission were taken into account explicitly. The dose distribution was calculated by convolving the in-air fluence distribution with an experimentally determined pencil-beam kernel. The results were compared with measurements using a diode array and passing rates with 2%/2 mm and 3%/3 mm criteria were reported. It was found that the two-source model achieved the best agreement on head scatter factor calculation. The three-source model and single-source model underestimated head scatter factors for certain symmetric rectangular fields and asymmetric fields, but similar good agreement could be achieved when monitor back scatter effect was incorporated explicitly. All the three source models resulted in comparable average passing rates (>97%) when the 3%/3 mm criterion was selected. The calculation with the single-source model and two-source model was slightly faster than the three-source model due to their simplicity.

  15. MR angiography in the diagnosis of tumors in the head and neck

    International Nuclear Information System (INIS)

    Vogl, T.J.; Balzer, J.O.; Juergens, M.; Lissner, J.; Grevers, G.

    1992-01-01

    40 normal individuals and 153 patients with lesions in the head and neck were examined by conventional imaging methods and by means of MR angiography (1.5 Tesla Magnetome). The problems to be solved concerned the ralationship between tumors and vessels and vascular anomalies and abnormalities at the skull base (56 cases), the facial skeleton (62 cases) and the neck (35 cases). Digital subtraction angiography was performed in 54 patients and the findings corelated with MR angiography. Optimal results were obtained by using a FISP 3D sequence; in this way arterial structures could be rendered reproducibly down to a diameter of 2 mm. The venous system in the head and neck was best shown by a FLASH 2D sequence. Correlation with arterial DSA showed high accuracy of MR angiography (91%) concerning displacement of vessels, the topography and the recognition of vascular occlusions. Our results indicate that MR angiography is a rapid and reliable procedure for evaluating the arterial and venous changes due to tumors in the head and neck region. (orig.) [de

  16. A simplified hydrodynamic model of hydrogen flame propagation in reactor vessels

    International Nuclear Information System (INIS)

    Baer, M.; Ratzel, A.

    1983-01-01

    A hydrodynamic model for hydrogen flame propagation in reactor geometries is presented. This model is consistent with the theory of slow combustion in which the gasdynamic field equations are treated in the limit of small Mach numbers. To the lowest order, pressure is spatially uniform. The flame is treated as a density and entropy discontinuity which propagates at prescribed burning velocities, corresponding to laminar or turbulent flames. Radiation cooling of the burned combustion gases and possible preheating of the unburned gases during propagation of the flame is included using a molecular gas-band thermal radiation model. Application of this model has been developed for 1-D variable area flame propagation. Multidimensional effects induced by hydrodynamics and buoyancy are introduced as a correction to the burn velocity (which reflects a modification of planar flame surface to a distorted surface) using experimentally measured pressure-rise time data for hydrogen/air deflagrations in cylindrical vessels. This semianalytical model of flame propagation reduces to a set of ordinary differential equations which describes the temporal variations of vessel pressure, burned volume and gas entropy. The thermodynamic state of the burned gas immediately following the flame is determined using an isobaric Hugoniot relationship. At other locations the burned gas thermodynamic states are determined using a Lagrangian particle tracking method. Results of a computer code using the method are presented

  17. Reactor pressure vessel embrittlement: Insights from neural network modelling

    Science.gov (United States)

    Mathew, J.; Parfitt, D.; Wilford, K.; Riddle, N.; Alamaniotis, M.; Chroneos, A.; Fitzpatrick, M. E.

    2018-04-01

    Irradiation embrittlement of steel pressure vessels is an important consideration for the operation of current and future light water nuclear reactors. In this study we employ an ensemble of artificial neural networks in order to provide predictions of the embrittlement using two literature datasets, one based on US surveillance data and the second from the IVAR experiment. We use these networks to examine trends with input variables and to assess various literature models including compositional effects and the role of flux and temperature. Overall, the networks agree with the existing literature models and we comment on their more general use in predicting irradiation embrittlement.

  18. Conjugate heat transfer analysis for in-vessel retention with external reactor vessel cooling

    International Nuclear Information System (INIS)

    Park, Jong-Woon; Bae, Jae-ho; Song, Hyuk-Jin

    2016-01-01

    Highlights: • A conjugate heat transfer analysis method is applied for in-vessel corium retention. • 3D heat diffusion has a formidable effect in alleviating focusing heat load from metallic layer. • The focusing heat load is decreased by about 2.5 times on the external surface. - Abstract: A conjugate heat transfer analysis method for the thermal integrity of a reactor vessel under external reactor vessel cooling conditions is developed to resolve light metal layer focusing effect issue for in-vessel retention. The method calculates steady-state three-dimensional temperature distribution of a reactor vessel using coupled conjugate heat transfer between in-vessel three-layered stratified corium (metallic pool, oxide pool and heavy metal and polar-angle dependent boiling heat transfer at the outer surface of a reactor vessel). The three-layer corium heat transfer model is utilizing lumped-parameter thermal-resistance circuit method. For the ex-vessel boiling boundary conditions, nucleate, transition and film boiling are considered. The thermal integrity of a reactor vessel is addressed in terms of heat flux at the outer-most nodes of the vessel and remaining thickness profile. The vessel three-dimensional heat conduction is validated against a commercial code. It is found that even though the internal heat flux from the metal layer goes far beyond critical heat flux (CHF) the heat flux from the outermost nodes of the vessel may be maintained below CHF due to massive vessel heat diffusion. The heat diffusion throughout the vessel is more pronounced for relatively low heat generation rate in an oxide pool. Parametric calculations are performed considering thermal conditions such as peak heat flux from a light metal layer, heat generation in an oxide pool and external boiling conditions. The major finding is that the most crucial factor for success of in-vessel retention is not the mass of the molten light metal above the oxide pool but the heat generation rate

  19. CFD Modeling of a Multiphase Gravity Separator Vessel

    KAUST Repository

    Narayan, Gautham

    2017-05-23

    The poster highlights a CFD study that incorporates a combined Eulerian multi-fluid multiphase and a Population Balance Model (PBM) to study the flow inside a typical multiphase gravity separator vessel (GSV) found in oil and gas industry. The simulations were performed using Ansys Fluent CFD package running on KAUST supercomputer, Shaheen. Also, a highlight of a scalability study is presented. The effect of I/O bottlenecks and using Hierarchical Data Format (HDF5) for collective and independent parallel reading of case file is presented. This work is an outcome of a research collaboration on an Aramco project on Shaheen.

  20. CFD Modeling of a Multiphase Gravity Separator Vessel

    KAUST Repository

    Narayan, Gautham; Khurram, Rooh Ul Amin; Elsaadawy, Ehab

    2017-01-01

    The poster highlights a CFD study that incorporates a combined Eulerian multi-fluid multiphase and a Population Balance Model (PBM) to study the flow inside a typical multiphase gravity separator vessel (GSV) found in oil and gas industry. The simulations were performed using Ansys Fluent CFD package running on KAUST supercomputer, Shaheen. Also, a highlight of a scalability study is presented. The effect of I/O bottlenecks and using Hierarchical Data Format (HDF5) for collective and independent parallel reading of case file is presented. This work is an outcome of a research collaboration on an Aramco project on Shaheen.

  1. A Constraint Programming Model for Fast Optimal Stowage of Container Vessel Bays

    DEFF Research Database (Denmark)

    Delgado-Ortegon, Alberto; Jensen, Rune Møller; Janstrup, Kira

    2012-01-01

    Container vessel stowage planning is a hard combinatorial optimization problem with both high economic and environmental impact. We have developed an approach that often is able to generate near-optimal plans for large container vessels within a few minutes. It decomposes the problem into a master...... planning phase that distributes the containers to bay sections and a slot planning phase that assigns containers of each bay section to slots. In this paper, we focus on the slot planning phase of this approach and present a constraint programming and integer programming model for stowing a set...... of containers in a single bay section. This so-called slot planning problem is NP-hard and often involves stowing several hundred containers. Using state-of-the-art constraint solvers and modeling techniques, however, we were able to solve 90% of 236 real instances from our industrial collaborator to optimality...

  2. Simplified methods to the complete thermal and mechanical behavior of a pressure vessel during a severe accident

    International Nuclear Information System (INIS)

    Dupas, P.; Schneiter, J.R.

    1996-01-01

    EDF has developed a software package of simplified methods (proprietary ones or from literature) in order to study the thermal and mechanical behavior of a PWR pressure vessel during a severe accident involving a corium localization in the vessel lower head. Using a part of this package, the authors can evaluate for instance successively: the heat flux at the inner surface of the vessel (conductive or convective pool of corium); the thermal exchange coefficient between the vessel and the outside (dry pit or flooded pit, watertight thermal insulation or not); the complete thermal evolution of the vessel (temperature profile, melting); the possible global plastic failure of the vessel; the creep behavior in the thickness of the vessel. These simplified methods are a cost effective alternative to finite element calculations which are yet used to validate the previous methods, waiting for experimental results to come

  3. Study on prediction model of irradiation embrittlement for reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Wang Rongshan; Xu Chaoliang; Huang Ping; Liu Xiangbing; Ren Ai; Chen Jun; Li Chengliang

    2014-01-01

    The study on prediction model of irradiation embrittlement for reactor pres- sure vessel (RPV) steel is an important method for long term operation. According to the deep analysis of the previous prediction models developed worldwide, the drawbacks of these models were given and a new irradiation embrittlement prediction model PMIE-2012 was developed. A corresponding reliability assessment was carried out by irradiation surveillance data. The assessment results show that the PMIE-2012 have a high reliability and accuracy on irradiation embrittlement prediction. (authors)

  4. Simplified methods applied to the complete thermal and mechanical behaviour of a pressure vessel during a severe accident

    International Nuclear Information System (INIS)

    Dupas, P.

    1996-01-01

    EDF has developed a software package of simplified methods (proprietary ones from literature) in order to study the thermal and mechanical behaviour of a PWR pressure vessel during a severe accident involving a corium localization in the vessel lower head. Using a part of this package, we can evaluate for instance successively: the heat flux at the inner surface of the vessel (conductive or convective pool of corium); the thermal exchange coefficient between the vessel and the outside (dry pit or flooded pit, watertight thermal insulation or not); the complete thermal evolution of the vessel (temperature profile, melting); the possible global plastic failure of the vessel; the creep behaviour in the vessel. These simplified methods are low cost alternative to finite element calculations which are yet used to validate the previous methods, waiting for experimental results to come. (authors)

  5. A scaling law for the local CHF on the external bottom side of a fully submerged reactor vessel

    International Nuclear Information System (INIS)

    Cheung, F.B.; Haddad, K.H.; Liu, Y.C.

    1997-01-01

    A scaling law for estimating the local critical heat flux on the outer surface of a heated hemispherical vessel that is fully submerged in water has been developed from the results of an advanced hydrodynamic CHF model for pool boiling on a downward facing curved heating surface. The scaling law accounts for the effects of the size of the vessel, the level of liquid subcooling, the intrinsic properties of the fluid, and the spatial variation of the local critical heat flux along the heating surface. It is found that for vessels with diameters considerably larger than the characteristic size of the vapor masses, the size effect on the local critical heat flux is limited almost entirely to the effect of subcooling associated with the local liquid head. When the subcooling effect is accounted for separately, the local CHF limit is nearly independent of the vessel size. Based upon the scaling law developed in this work, it is possible to merge, within the experimental uncertainties, all the available local CHF data obtained for various vessel sizes under both saturated and subcooled boiling conditions into a single curve. Applications of the scaling law to commercial-size vessels have been made for various system pressures and water levels above the heated vessel. Over the range of conditions explored in this study, the local CHF limit is found to increase by a factor of two or more from the bottom center to the upper edge of the vessel. Meanwhile, the critical heat flux at a given angular position of the heated vessel is also found to increase appreciably with the system pressure and the water level

  6. Analytical model for shear strength of end slabs of prestressed concrete nuclear reactor vessels

    International Nuclear Information System (INIS)

    Abdulrahman, H.O.; Sozen, M.A.; Schnobrich, W.C.

    1979-04-01

    The results are presented of an investigation of the behavior and strength of flat end slabs of cylindrical prestressed concrete nuclear reactor vessels. The investigation included tests of ten small-scale pressure vessels and development of a nonlinear finite-element model to simulate the deformation response and strength of the end slabs. Because earlier experimental studies had shown that the flexural strength of the end slab could be calculated using intelligible procedures, the emphasis of this investigation was on shear strength

  7. Experiments on melt dispersion with lateral failure in the bottom head of the pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, L.; Gargallo, M. [Forschungszentrum Karlsruhe, Institut fur Kern-und Energietechnik, Karlsruhe (Germany)

    2001-07-01

    Melt dispersion experiments with lateral failure in the bottom head were carried out in a 1:18 scaled annular cavity design under low pressure conditions. Water and a bismuth alloy were used as melt simulant material and nitrogen as driving gas. With lateral breaches the liquid height in the lower head relative to the upper and lower edge of the breach is an additional parameter for the dispersion process. Shifting the break from the central position towards the side of the lower head leads to smaller melt dispersion, and a larger breach size does not necessarily lead to a larger dispersed melt fraction. (author)

  8. Tightening unit EZ 250 for VVER 1000 type reactor pressure vessel head flange joints

    International Nuclear Information System (INIS)

    Ruchar, Miloslav; Nadenik, Tomas; Kroj, Ludek

    2010-01-01

    The programme of flange joints tightening by seals made of expanded graphite for VVER 440 and VVER 1000 reactor head flange joints is highlighted, and tightening units of row EZ 650 and EZ 650 TK and KNI for VVER 440 reactor head flange joints and EZ 250 tightening unit for VVER 1000 reactor head flange joints are described in detail. The main advantages of electronically controlled tightening units include: Precise and uniform compression of the gasket during the tightening procedure; Automated solution to the graphite relaxing problem during tightening; Possibility of diagnosis of the thread status of the joints being tightened; Alleviation of operator's tough work; Shorter time of tensioning associated with a lower collective doses; Quick preparation of tightening procedure report from the data measured; Calibration renewal is possible in advance at time of unit storage without the need to place it on a real joint. (P.A.)

  9. BrainK for Structural Image Processing: Creating Electrical Models of the Human Head

    Directory of Open Access Journals (Sweden)

    Kai Li

    2016-01-01

    Full Text Available BrainK is a set of automated procedures for characterizing the tissues of the human head from MRI, CT, and photogrammetry images. The tissue segmentation and cortical surface extraction support the primary goal of modeling the propagation of electrical currents through head tissues with a finite difference model (FDM or finite element model (FEM created from the BrainK geometries. The electrical head model is necessary for accurate source localization of dense array electroencephalographic (dEEG measures from head surface electrodes. It is also necessary for accurate targeting of cerebral structures with transcranial current injection from those surface electrodes. BrainK must achieve five major tasks: image segmentation, registration of the MRI, CT, and sensor photogrammetry images, cortical surface reconstruction, dipole tessellation of the cortical surface, and Talairach transformation. We describe the approach to each task, and we compare the accuracies for the key tasks of tissue segmentation and cortical surface extraction in relation to existing research tools (FreeSurfer, FSL, SPM, and BrainVisa. BrainK achieves good accuracy with minimal or no user intervention, it deals well with poor quality MR images and tissue abnormalities, and it provides improved computational efficiency over existing research packages.

  10. In-vessel core melt retention by RPV external cooling for high power PWR. MAAP 4 analysis on a LBLOCA scenario without SI

    International Nuclear Information System (INIS)

    Cognet, C.; Gandrille, P.

    1999-01-01

    In-, ex-vessel reflooding or both simultaneously can be envisaged as Accident Management Measures to stop a Severe Accident (SA) in vessel. This paper addresses the possibility of in-vessel core melt retention by RPV external flooding for a high power PWR (4250 MWth). The reactor vessel is assumed to have no lower head penetration and thermal insulation is neglected. The effects of external cooling of high power density debris, where the margin for such a strategy is low, are investigated with the MAAP4 code. MAAP4 code is used to verify the system capability to flood the reactor pit and to predict simultaneously the corium relocation into the lower head with the thermal and mechanical response of the RPV in transient conditions. The corium pool cooling and holding in the RPV lower head is analysed. Attention is paid to the internal heat exchanges between corium components. This paper focuses particularly the heat transfer between oxidic and metallic phases as well as between the molten metallic phase and the RPV wall of utmost importance for challenging the RPV integrity in vicinity of the metallic phase. The metal segregation has a decisive influence upon the attack of the vessel wall due to a very strong peaking of the lateral flux ('focusing effect'). Thus, the dynamics of the formation of the metallic layer characterized by a growing inventory of steel, both from a partial vessel ablation and the degradation of internals steel structures by the radiative heat flux from the debris, is displayed. The analysed sequence is a surge line rupture near the hot leg (LBLOCA) leading to the fastest accident progression

  11. A Model of Self-Organizing Head-Centered Visual Responses in Primate Parietal Areas

    Science.gov (United States)

    Mender, Bedeho M. W.; Stringer, Simon M.

    2013-01-01

    We present a hypothesis for how head-centered visual representations in primate parietal areas could self-organize through visually-guided learning, and test this hypothesis using a neural network model. The model consists of a competitive output layer of neurons that receives afferent synaptic connections from a population of input neurons with eye position gain modulated retinal receptive fields. The synaptic connections in the model are trained with an associative trace learning rule which has the effect of encouraging output neurons to learn to respond to subsets of input patterns that tend to occur close together in time. This network architecture and synaptic learning rule is hypothesized to promote the development of head-centered output neurons during periods of time when the head remains fixed while the eyes move. This hypothesis is demonstrated to be feasible, and each of the core model components described is tested and found to be individually necessary for successful self-organization. PMID:24349064

  12. Modeling heading and path perception from optic flow in the case of independently moving objects

    Science.gov (United States)

    Raudies, Florian; Neumann, Heiko

    2013-01-01

    Humans are usually accurate when estimating heading or path from optic flow, even in the presence of independently moving objects (IMOs) in an otherwise rigid scene. To invoke significant biases in perceived heading, IMOs have to be large and obscure the focus of expansion (FOE) in the image plane, which is the point of approach. For the estimation of path during curvilinear self-motion no significant biases were found in the presence of IMOs. What makes humans robust in their estimation of heading or path using optic flow? We derive analytical models of optic flow for linear and curvilinear self-motion using geometric scene models. Heading biases of a linear least squares method, which builds upon these analytical models, are large, larger than those reported for humans. This motivated us to study segmentation cues that are available from optic flow. We derive models of accretion/deletion, expansion/contraction, acceleration/deceleration, local spatial curvature, and local temporal curvature, to be used as cues to segment an IMO from the background. Integrating these segmentation cues into our method of estimating heading or path now explains human psychophysical data and extends, as well as unifies, previous investigations. Our analysis suggests that various cues available from optic flow help to segment IMOs and, thus, make humans' heading and path perception robust in the presence of such IMOs. PMID:23554589

  13. Modeling Heading and Path Perception from Optic Flow in the Case of Independently Moving Objects

    Directory of Open Access Journals (Sweden)

    Florian eRaudies

    2013-04-01

    Full Text Available Humans are usually accurate when estimating heading or path from optic flow, even in the presence of independently moving objects (IMO in an otherwise rigid scene. To invoke significant biases in perceived heading, IMOs have to be large and obscure the focus of expansion (FOE in the image plane, which is the point of approach. For the estimation of path during curvilinear self-motion no significant biases were found in the presence of IMOs. What makes humans robust in their estimation of heading or path using optic flow? We derive analytical models of optic flow for linear and curvilinear self-motion using geometric scene models. Heading biases of a linear least squares method, which builds upon these analytical models, are large, larger than those reported for humans. This motivated us to study segmentation cues that are available from optic flow. We derive models of accretion / deletion, expansion / contraction, acceleration / deceleration, local spatial curvature, and local temporal curvature, to be used as cues to segment an IMO from the background. Integrating these segmentation cues into our method of estimating heading or path now explains human psychophysical data and extends, as well as unifies, previous investigations. Our analysis suggests that various cues available from optic flow help to segment IMOs and, thus, make humans’ heading and path perception robust in the presence of such IMOs.

  14. A mathematical model for cost of maritime transport. Application to competitiveness of nuclear vessels

    International Nuclear Information System (INIS)

    Dorval, C.

    1966-05-01

    In studying the competitiveness of a nuclear merchant vessel, economic assessments in terms of figures were discarded in favor of a simplified model, which gives a clearer idea of the mechanism of the comparison between alternative vessels and the particular influence of each parameter. An expression is formulated for the unit cost per ton carried over a given distance as a function of the variables (speed and deadweight tonnage) and is used to determine the optima for conventional and nuclear vessels. To represent the freight market involved in the optimization studies, and thus in the competitiveness computation, two cases are taken into account: the tonnage to be carried annually is limited, and the tonnage to be carried annually is not limited. In both cases the optima are calculated and compared for a conventional and a nuclear vessel. Competitiveness curves are plotted as a function of the ratios of nuclear and conventional fuel costs and nuclear and conventional marginal power costs. These curves express the limiting values of the above two ratios for which the transport costs of the nuclear and conventional vessels are equal. The competitiveness curves vary considerably according to the hypothesis adopted for the freight market and the limit of tonnage carried annually. (author) [fr

  15. Segmentation of vessels cluttered with cells using a physics based model.

    Science.gov (United States)

    Schmugge, Stephen J; Keller, Steve; Nguyen, Nhat; Souvenir, Richard; Huynh, Toan; Clemens, Mark; Shin, Min C

    2008-01-01

    Segmentation of vessels in biomedical images is important as it can provide insight into analysis of vascular morphology, topology and is required for kinetic analysis of flow velocity and vessel permeability. Intravital microscopy is a powerful tool as it enables in vivo imaging of both vasculature and circulating cells. However, the analysis of vasculature in those images is difficult due to the presence of cells and their image gradient. In this paper, we provide a novel method of segmenting vessels with a high level of cell related clutter. A set of virtual point pairs ("vessel probes") are moved reacting to forces including Vessel Vector Flow (VVF) and Vessel Boundary Vector Flow (VBVF) forces. Incorporating the cell detection, the VVF force attracts the probes toward the vessel, while the VBVF force attracts the virtual points of the probes to localize the vessel boundary without being distracted by the image features of the cells. The vessel probes are moved according to Newtonian Physics reacting to the net of forces applied on them. We demonstrate the results on a set of five real in vivo images of liver vasculature cluttered by white blood cells. When compared against the ground truth prepared by the technician, the Root Mean Squared Error (RMSE) of segmentation with VVF and VBVF was 55% lower than the method without VVF and VBVF.

  16. 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

  17. In-vessel core debris retention through external flooding of the reactor pressure vessel. State-of-the-art report

    Energy Technology Data Exchange (ETDEWEB)

    Heel, A.M.J.M. van

    1995-07-01

    An overview of the state-of-the-art knowledge on the ex-vessel flooding accident management strategy for severe accidents in a NPP has been given. The feasibility has been discussed, as well as the in- and ex-vessel phenomena, which influence the structural integrity of the vessel. Finally, some computer codes with the ability to model the phenomena involved in ex-vessel flooding have been discussed. (orig./HP).

  18. In-vessel core debris retention through external flooding of the reactor pressure vessel. State-of-the-art report

    International Nuclear Information System (INIS)

    Heel, A.M.J.M. van.

    1995-07-01

    An overview of the state-of-the-art knowledge on the ex-vessel flooding accident management strategy for severe accidents in a NPP has been given. The feasibility has been discussed, as well as the in- and ex-vessel phenomena, which influence the structural integrity of the vessel. Finally, some computer codes with the ability to model the phenomena involved in ex-vessel flooding have been discussed. (orig./HP)

  19. Heading-vector navigation based on head-direction cells and path integration.

    Science.gov (United States)

    Kubie, John L; Fenton, André A

    2009-05-01

    Insect navigation is guided by heading vectors that are computed by path integration. Mammalian navigation models, on the other hand, are typically based on map-like place representations provided by hippocampal place cells. Such models compute optimal routes as a continuous series of locations that connect the current location to a goal. We propose a "heading-vector" model in which head-direction cells or their derivatives serve both as key elements in constructing the optimal route and as the straight-line guidance during route execution. The model is based on a memory structure termed the "shortcut matrix," which is constructed during the initial exploration of an environment when a set of shortcut vectors between sequential pairs of visited waypoint locations is stored. A mechanism is proposed for calculating and storing these vectors that relies on a hypothesized cell type termed an "accumulating head-direction cell." Following exploration, shortcut vectors connecting all pairs of waypoint locations are computed by vector arithmetic and stored in the shortcut matrix. On re-entry, when local view or place representations query the shortcut matrix with a current waypoint and goal, a shortcut trajectory is retrieved. Since the trajectory direction is in head-direction compass coordinates, navigation is accomplished by tracking the firing of head-direction cells that are tuned to the heading angle. Section 1 of the manuscript describes the properties of accumulating head-direction cells. It then shows how accumulating head-direction cells can store local vectors and perform vector arithmetic to perform path-integration-based homing. Section 2 describes the construction and use of the shortcut matrix for computing direct paths between any pair of locations that have been registered in the shortcut matrix. In the discussion, we analyze the advantages of heading-based navigation over map-based navigation. Finally, we survey behavioral evidence that nonhippocampal

  20. 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)

  1. Development of Head Injury Assessment Reference Values Based on NASA Injury Modeling

    Science.gov (United States)

    Somers, Jeffrey T.; Melvin, John W.; Tabiei, Ala; Lawrence, Charles; Ploutz-Snyder, Robert; Granderson, Bradley; Feiveson, Alan; Gernhardt, Michael; Patalak, John

    2011-01-01

    NASA is developing a new capsule-based, crewed vehicle that will land in the ocean, and the space agency desires to reduce the risk of injury from impact during these landings. Because landing impact occurs for each flight and the crew might need to perform egress tasks, current injury assessment reference values (IARV) were deemed insufficient. Because NASCAR occupant restraint systems are more effective than the systems used to determine the current IARVs and are similar to NASA s proposed restraint system, an analysis of NASCAR impacts was performed to develop new IARVs that may be more relevant to NASA s context of vehicle landing operations. Head IARVs associated with race car impacts were investigated by completing a detailed analysis of all of the 2002-2008 NASCAR impact data. Specific inclusion and exclusion criteria were used to select 4071 impacts from the 4015 recorder files provided (each file could contain multiple impact events). Of the 4071 accepted impacts, 274 were selected for numerical simulation using a custom NASCAR restraint system and Humanetics Hybrid-III 50th percentile numerical dummy model in LS-DYNA. Injury had occurred in 32 of the 274 selected impacts, and 27 of those injuries involved the head. A majority of the head injuries were mild concussions with or without brief loss of consciousness. The 242 non-injury impacts were randomly selected and representative of the range of crash dynamics present in the total set of 4071 impacts. Head dynamics data (head translational acceleration, translational change in velocity, rotational acceleration, rotational velocity, HIC-15, HIC-36, and the Head 3ms clip) were filtered according to SAE J211 specifications and then transformed to a log scale. The probability of head injury was estimated using a separate logistic regression analysis for each log-transformed predictor candidate. Using the log transformation constrains the estimated probability of injury to become negligible as IARVs approach

  2. 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)

  3. Simulant melt experiments on performance of the in-vessel core catcher

    International Nuclear Information System (INIS)

    Kang, Kyoung-Ho; Park, Rae-Joon; Kim, Sang-Baik; Suh, K.Y.; Cheung, F.B.; Rempe, J.L.

    2007-01-01

    In order to enhance the feasibility of in-vessel retention (IVR) of molten core material during a severe accident for high-power reactors, an in-vessel core catcher (IVCC) was designed and evaluated as part of a joint United States-Korean International Nuclear Energy Research Initiative (INERI). The proposed IVCC is expected to increase the thermal margin for success of IVR by providing an 'engineered gap' for heat transfer from materials that relocate during a severe accident and potentially serving as a sacrificial material under a severe accident. In this study, LAVA-GAP experiments were performed to investigate the thermal and mechanical performance of the IVCC using the alumina melt as simulant. The LAVA-GAP experiments aim to examine the feasibility and sustainability of the IVCC under the various test conditions using 1/8th scale hemispherical test sections. As a feasibility test of the proposed IVCC in this INERI project, the effects of IVCC base steel materials, internal coating materials, and gap size between the IVCC and the vessel lower head were examined. The test results indicated that the internally coated IVCC has high thermal performance compared with the uncoated IVCC. In terms of integrity of the base steel, carbon steel is superior to stainless steel and the effect of bond coat is found to be trivial for the tests performed in this study. The thermal load is mitigated via boiling heat removal in the gap between the IVCC and the vessel lower head. The current test results imply that gaps less than 10 mm are not enough to guarantee effective cooling induced by water ingression and steam venting there through. Selection of endurable material and pertinent gap size is needed to implement the proposed IVCC concept into advanced reactor designs

  4. Navigation in head and neck oncological surgery: an emerging concept.

    Science.gov (United States)

    Gangloff, P; Mastronicola, R; Cortese, S; Phulpin, B; Sergeant, C; Guillemin, F; Eluecque, H; Perrot, C; Dolivet, G

    2011-01-01

    Navigation surgery, initially applied in rhinology, neurosurgery and orthopaedic cases, has been developed over the last twenty years. Surgery based on computed tomography data has become increasingly important in the head and neck region. The technique for hardware fusion between RMI and computed tomography is also becoming more useful. We use such device since 2006 in head and neck carcinologic situation. Navigation allows control of the resection in order to avoid and protect the precise anatomical structures (vessels and nerves). It also guides biopsy and radiofrequency. Therefore, quality of life is much more increased and morbidity is decreased for these patients who undergo major and mutilating head and neck surgery. Here we report the results of 33 navigation procedures performed for 31 patients in our institution.

  5. MATHEMATICAL MODEL OF TRIAXIAL MULTIMODE ATTITUDE AND HEADING REFERENCE SYSTEM

    Directory of Open Access Journals (Sweden)

    Olha Sushchenko

    2017-07-01

    Full Text Available Purpose: The paper deals with the mathematical description of the gimballed attitude and heading reference systems, which can be applied in design of strategic precision navigation systems. The main goal is to created mathematical description taking into consideration the necessity to use different navigations operating modes of this class of navigation systems. To provide the high accuracy the indirect control is used when the position of the gimballed platform is controlled by signals of gyroscopic devices, which are corrected using accelerometer’s signals. Methods: To solve the given problem the methods of the classical theoretical mechanics, gyro theory, and inertial navigation are used. Results: The full mathematical model of the gimballed attitude and heading reference system is derived including descriptions of different operating modes. The mathematical models of the system Expressions for control and correction moments in the different modes are represented. The simulation results are given. Conclusions: The represented results prove efficiency of the proposed models. Developed mathematical models can be useful for design of navigation systems of the wide class of moving vehicles.

  6. Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

    Energy Technology Data Exchange (ETDEWEB)

    Urbonavičius, E., E-mail: Egidijus.Urbonavicius@lei.lt; Povilaitis, M., E-mail: Mantas.Povilaitis@lei.lt; Kontautas, A., E-mail: Aurimas.Kontautas@lei.lt

    2015-11-15

    Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

  7. Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

    International Nuclear Information System (INIS)

    Urbonavičius, E.; Povilaitis, M.; Kontautas, A.

    2015-01-01

    Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

  8. Matching the results of a theoretical model with failure rates obtained from a population of non-nuclear pressure vessels

    International Nuclear Information System (INIS)

    Harrop, L.P.

    1982-02-01

    Failure rates for non-nuclear pressure vessel populations are often regarded as showing a decrease with time. Empirical evidence can be cited which supports this view. On the other hand theoretical predictions of PWR type reactor pressure vessel failure rates have shown an increasing failure rate with time. It is shown that these two situations are not necessarily incompatible. If adjustments are made to the input data of the theoretical model to treat a non-nuclear pressure vessel population, the model can produce a failure rate which decreases with time. These adjustments are explained and the results obtained are shown. (author)

  9. Mechanical modelling of a structural performance of a pressure vessel submitted to the creep phenomenon

    International Nuclear Information System (INIS)

    Taroco, E.; Feijoo, R.A.; Monteiro, Edson; Freire, J.L.F.; Bevilacqua, L.; Miranda, P.E.V. de; Silveira, T.L. da

    1982-01-01

    A pressure vessel is analized using different mechanical models for the creep phenomenon. The numerical results obtained through these models enable us to recommend on the way verifications of creep damage accumulation is structures should be made. (Author) [pt

  10. Vessel eddy current characteristics in SST-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jana, Subrata; Pradhan, Subrata, E-mail: pradhan@ipr.res.in; Dhongde, Jasraj; Masand, Harish

    2016-11-15

    Highlights: • Eddy current distribution in the SST-1 vacuum vessel. • Circuit model analysis of eddy current. • A comparison of the field lines with and without the plasma column in identical conditions. • The influence of eddy current in magnetic NULL dynamics. - Abstract: Eddy current distribution in the vacuum vessel of the Steady state superconducting (SST-1) tokamak has been determined from the experimental data obtained using an array of internal voltage loops (flux loop) installed inside the vacuum vessel. A simple circuit model has been employed. The model takes into account the geometric and constructional features of SST-1 vacuum vessel. SST-1 vacuum vessel is a modified ‘D’ shaped vessel having major axis of 1.285 m and minor axis of 0.81 m and has been manufactured from non-magnetic stainless steel. The Plasma facing components installed inside the vacuum vessel are graphite blocks mounted on Copper Chromium Zirconium (CuCrZr) heat sink plates on inconel supports. During discharge of the central solenoid, eddy currents get generated in the vacuum vessel and passive supports on it. These eddy currents influence the early magnetic NULL dynamics and plasma break-down and start-up characteristics. The computed results obtained from the model have been benchmarked against experimental data obtained in large number of SST-1 plasma shots. The results are in good agreement. Once bench marked, the calculated eddy current based on flux loop signal and circuit equation model has been extended to the reconstruction of the overall B- field contours of SST-1 tokamak in the vessel region. A comparison of the field lines with and without the plasma column in identical conditions of the central solenoid and equilibrium field profiles has also been done with an aim to quantify the diagnostics responses in vacuum shots.

  11. Lower head integrity under steam explosion loads

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, T.G.; Yuen, W.W.; Angelini, S.; Freeman, K.; Chen, X.; Salmassi, T. [Center for Risk Studies and Safety, Univ. of California, Santa Barbara, CA (United States); Sienicki, J.J.

    1998-01-01

    Lower head integrity under steam explosion loads in an AP600-like reactor design is considered. The assessment is the second part of an evaluation of the in-vessel retention idea as a severe accident management concept, the first part (DOE/ID-10460) dealing with thermal loads. The assessment is conducted in terms of the Risk Oriented Accident Analysis Methodology (ROAAM), and includes the comprehensive evaluation of all relevant severe accident scenarios, melt conditions and timing of release from the core region, fully 3D mixing and explosion wave dynamics, and lower head fragility under local, dynamic loading. All of these factors and brought together in a ROAAM Probabilistic Framework to evaluate failure likelihood. The conclusion is that failure is `physically unreasonable`. (author)

  12. Preliminary analysis of a 1:4 scale prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Dameron, R.A.; Rashid, Y.R.; Luk, V.K.; Hessheimer, M.F.

    1997-01-01

    Sandia National Laboratories is conducting a research program to investigate the integrity of nuclear containment structures. As part of the program Sandia will construct an instrumented 1:4 scale model of a prestressed concrete containment vessel (PCCV) for pressurized water reactors (PWR), which will be pressure tested up to its ultimate capacity. One of the key program objectives is to develop validated methods to predict the structural performance of containment vessels when subjected to beyond design basis loadings. Analytical prediction of structural performance requires a stepwise, systematic approach that addresses all potential failure modes. The analysis effort includes two and three-dimensional nonlinear finite element analyses of the PCCV test model to evaluate its structural performance under very high internal pressurization. Such analyses have been performed using the nonlinear concrete constitutive model, ANACAP-U, in conjunction with the ABAQUS general purpose finite element code. The analysis effort is carried out in three phases: preliminary analysis; pretest prediction; and post-test data interpretation and analysis evaluation. The preliminary analysis phase serves to provide instrumentation support and identify candidate failure modes. The associated tasks include the preliminary prediction of failure pressure and probable failure locations and the development of models to be used in the detailed failure analyses. This paper describes the modeling approaches and some of the results obtained in the first phase of the analysis effort

  13. Immunocompromised and immunocompetent mouse models for head and neck squamous cell carcinoma

    Directory of Open Access Journals (Sweden)

    Lei ZG

    2016-01-01

    Full Text Available Zhen-ge Lei,1,* Xiao-hua Ren,2,* Sha-sha Wang,3 Xin-hua Liang,3,4 Ya-ling Tang3,5 1Department of Oral and Maxillofacial Surgery, Stomatological Hospital Affiliated to Nanchang University, Nanchang, Jiangxi, 2Department of Stomatology, Sichuan Medical Science Academy and Sichuan Provincial People’s Hospital, 3State Key Laboratory of Oral Diseases, West China Hospital of Stomatology, Sichuan University, 4Department of Oral and Maxillofacial Surgery, West China College of Stomatology, Sichuan University, 5Department of Oral Pathology, West China Hospital of Stomatology, Sichuan University, Chengdu, Sichuan, People’s Republic of China *These authors contributed equally to this work Abstract: Mouse models can closely mimic human oral squamous epithelial carcinogenesis, greatly expand the in vivo research possibilities, and play a critical role in the development of diagnosis, monitoring, and treatment of head and neck squamous cell carcinoma. With the development of the recent research on the contribution of immunity/inflammation to cancer initiation and progression, mouse models have been divided into two categories, namely, immunocompromised and immunocompetent mouse models. And thus, this paper will review these two kinds of models applied in head and neck squamous cell carcinoma to provide a platform to understand the complicated histological, molecular, and genetic changes of oral squamous epithelial tumorigenesis. Keywords: head and neck squamous cell carcinoma, HNSCC, mouse models, immunocompromised models, immunocompetent models, transgenic models

  14. Comparisons with measured data of the simulated local core parameters by the coupled code ATHLET-BIPR-VVER applying a new enhanced model of the reactor pressure vessel

    International Nuclear Information System (INIS)

    Nikonov, S.; Pasichnyk, I.; Velkov, K.; Pautz, A.

    2011-01-01

    The paper describes the performed comparisons of measured and simulated local core data based on the OECD/NEA Benchmark on Kalinin-3 NPP: 'Switching off of one of the four operating main circulation pumps at nominal reactor power'. The local measurements of in core self-powered neutron detectors (SPND) in 64 fuel assemblies on 7 axial levels are used for the comparisons of the assemblies axial power distributions and the thermocouples readings at 93 fuel assembly heads are applied for the fuel assembly coolant temperature comparisons. The analyses are done on the base of benchmark transient calculations performed with the coupled system code ATHLET/BIPR-VVER. In order to describe more realistically the fluid mixing phenomena in a reactor pressure vessel a new enhanced nodalization scheme is being developed. It could take into account asymmetric flow behaviour in the reactor pressure vessel structures like downcomer, reactor core inlet and outlet, control rods' guided tubes, support grids etc. For this purpose details of the core geometry are modelled. About 58000 control volumes and junctions are applied. Cross connection are used to describe the interaction between the fluid objects. The performed comparisons are of great interest because they show some advantages by performing coupled code production pseudo-3D analysis of NPPs applying the parallel thermo-hydraulic channel methodology (or 1D thermo-hydraulic system code modeling). (Authors)

  15. In-vessel coolability and retention of a core melt. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T. [California Univ., Santa Barbara, CA (United States). Center for Risk Studies and Safety

    1996-10-01

    The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided.

  16. In-vessel coolability and retention of a core melt. Volume 2

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T.

    1996-10-01

    The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided

  17. In-vessel coolability and retention of a core melt. Volume 1

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T.

    1996-10-01

    The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided

  18. Non-destructive and destructive examination of the retired North Anna 2 Reactor Pressure Vessel Head

    International Nuclear Information System (INIS)

    Ahluwalia, Kawaljit; Barnes, Robert; Rao, Gutti; Cattant, Francois; Peat, Noel

    2006-09-01

    Stress corrosion cracking of Alloy 600 and nickel-based weld materials has been the single biggest challenge facing the PWR industry. A fundamental and thorough knowledge was needed to properly explain this phenomenon and develop appropriate mitigation strategies. Non Destructive Examination (NDE) of the North Anna Unit 2 Reactor Vessel Head (RVH) during the 2002 fall outage identified widespread crack indications in the Alloy 600 CRDM penetrations and associated Alloy 182 and 82 J-groove attachment welds. When the Utility decided to replace the RVH, a rare opportunity was provided to the industry to undertake in-depth studies of representative defective CRDM penetrations from a retired RVH. Accordingly, the Materials Reliability Program, undertook a two-phase program on the retired North Anna 2 Alloy 600 RVH. The first phase involved selection and removal of six penetrations from the RVH and penetration decontamination, replication and laboratory NDE. The second phase consisted of a detailed destructive examination of penetration number 54. This paper provides a summary of work undertaken during this program. Criteria for selection of penetrations for removal and procedures used in removal of the penetrations are described. Extreme care was undertaken in decontamination of the penetrations to facilitate laboratory NDE. Penetration number 54 was then subjected to destructive examination to establish a correlation between NDE findings (from both field and laboratory inspections) and actual flaws. Additional objectives of the destructive examination included mechanistic assessment of defect formation and investigation of the annulus environment and wastage characterization. Data obtained from these studies is invaluable in validating safety assessment statements by developing the correlation between field NDE and actual defects. In addition, information gathered from non-destructive and destructive examinations is used to assess accuracy of the NDE techniques

  19. Hydrodynamic model of hydrogen-flame propagation in reactor vessels

    International Nuclear Information System (INIS)

    Baer, M.R.; Ratzel, A.C.

    1982-01-01

    A hydrodynamic model for hydrogen flame propagation in reactor geometries is presented. This model is consistent with the theory of slow combustion in which the gasdynamic field equations are treated in the limit of small Mach numbers. To the lowest order, pressure is spatially uniform. The flame is treated as a density and entropy discontinuity which propagates at prescribed burning velocities, corresponding to laminar or turbulent flames. Radiation cooling of the burned combustion gases and possible preheating of the unburned gases during propagation of the flame is included using a molecular gas-band thermal radiation model. Application of this model has been developed for 1-D variable area flame propagation. Multidimensional effects induced by hydrodynamics and buoyancy are introduced as a correction to the burn velocity (which reflects a modification of planar flame surface to a distorted surface) using experimentally measured pressure-rise time data for hydrogen/air deflagrations in cylindrical vessels

  20. 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.

  1. A wave propagation model of blood flow in large vessels using an approximate velocity profile function

    NARCIS (Netherlands)

    Bessems, D.; Rutten, M.C.M.; Vosse, van de F.N.

    2007-01-01

    Lumped-parameter models (zero-dimensional) and wave-propagation models (one-dimensional) for pressure and flow in large vessels, as well as fully three-dimensional fluid–structure interaction models for pressure and velocity, can contribute valuably to answering physiological and patho-physiological

  2. Can Predictive Modeling Identify Head and Neck Oncology Patients at Risk for Readmission?

    Science.gov (United States)

    Manning, Amy M; Casper, Keith A; Peter, Kay St; Wilson, Keith M; Mark, Jonathan R; Collar, Ryan M

    2018-05-01

    Objective Unplanned readmission within 30 days is a contributor to health care costs in the United States. The use of predictive modeling during hospitalization to identify patients at risk for readmission offers a novel approach to quality improvement and cost reduction. Study Design Two-phase study including retrospective analysis of prospectively collected data followed by prospective longitudinal study. Setting Tertiary academic medical center. Subjects and Methods Prospectively collected data for patients undergoing surgical treatment for head and neck cancer from January 2013 to January 2015 were used to build predictive models for readmission within 30 days of discharge using logistic regression, classification and regression tree (CART) analysis, and random forests. One model (logistic regression) was then placed prospectively into the discharge workflow from March 2016 to May 2016 to determine the model's ability to predict which patients would be readmitted within 30 days. Results In total, 174 admissions had descriptive data. Thirty-two were excluded due to incomplete data. Logistic regression, CART, and random forest predictive models were constructed using the remaining 142 admissions. When applied to 106 consecutive prospective head and neck oncology patients at the time of discharge, the logistic regression model predicted readmissions with a specificity of 94%, a sensitivity of 47%, a negative predictive value of 90%, and a positive predictive value of 62% (odds ratio, 14.9; 95% confidence interval, 4.02-55.45). Conclusion Prospectively collected head and neck cancer databases can be used to develop predictive models that can accurately predict which patients will be readmitted. This offers valuable support for quality improvement initiatives and readmission-related cost reduction in head and neck cancer care.

  3. Earley changes in cerebral blood flow following head exposure of rabbit at a mean absorbed dose of 1000 rads

    International Nuclear Information System (INIS)

    Dufour, Raymond.

    1978-10-01

    Observations were made on 5 animals after 12 head exposures. The local cerebral blood flow increased on the 1st hour and lasted during the following 15-20 h. Two particular responses were used to test cerebral vasotonicity: rate increases due to inhalation of CO 2 (5 p. cent CO 2 in air) and accompanying paradoxical sleep. Their decrease after irradiation was studied systematically by CO 2 tests whatever the rate level reached. These weaker increases were not the indication of an upper limit of the the rate, they also occurred when the rate fell back to its initial value. They appeared on the 2nd hour and lasted till the 48th hour. As a conclusion: head irradiation resulted in an early increase of cerebral blood flow lasting less than 24 h; the pattern of metabolic or myogenic regulation of intraparenchymatous vessels and neurogenic regulation of extraparenchymatous vessels would explain the observed phenomena better. The metabolic modifications of the irradiated tissue should result in vasodilation of intraparenchymatous vessels. The stimulating action of the neurovegetative centers would be shown by a decrease of the vessel responses to CO 2 at the level of the extraparenchymatous vessels [fr

  4. Revision of the fracture models in steels for nuclear pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Darwish, F A.I. [Pontificia Univ. Catolica do Rio de Janeiro (Brazil). Dept. de Ciencia dos Materiais e Metalurgia

    1981-01-01

    The variation of toughness with the temperature of steels used in the fabrication of nuclear pressure vessels is presented and discuted by mathematical models aiming to reach a critical value of stress or deformation at the moment of the fracture. The mathematical model considered are compatible with the fracture micromechanisms in action and they are capable of foreseeing the variations in the toughness from the mechanical properties evaluated in the tension test. The neutron irradiation effects in the toughness as well as in the variation of this toughness with the operating temperature are still described.

  5. Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel: Part 1

    International Nuclear Information System (INIS)

    Macbeth, R.V.; Trenberth, R.

    1987-12-01

    Modelling experiments have been done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel. Various liquids have been used to simulate molten core debris assumed to have fallen on to the vault floor from a breach at the bottom of the pressure vessel. High pressure air and helium have been used to simulate the discharge of steam and gas from the breach. The dispersion of liquid via the vault access shafts has been measured. Photographs have been taken of fluid flow patterns and velocity profiles have been obtained. The requirements for further experiments are indicated. (author)

  6. Construction of tomographic head model using sectioned photographic images of cadaver

    International Nuclear Information System (INIS)

    Lee, Choon Sik; Lee, Jai Ki; Park, Jin Seo; Chung, Min Suk

    2004-01-01

    Tomographic models are currently the most complete, developed and realistic models of the human anatomy. They have been used to estimate organ doses for diagnostic radiation examination and radiotherapy treatment planning, and radiation protection. The quality of original anatomic images is a key factor to build a quality tomographic model. Computed tomography (CT) and magnetic resonance imaging (MRI) scan, from which most of current tomographic models are constructed, have their inherent shortcomings. In this study, a tomographic model of Korean adult male head was constructed by using serially sectioned photographs of cadaver. The cadaver was embedded, frozen, serially sectioned and photographed by high resolution digital camera at 0.2 mm interval. The contours of organs and tissues in photographs were segmented by several trained anatomists. The 120 segmented images of head at 2mm interval were converted into binary files and ported into Monte Carlo code to perform an example calculation of organ dose. Whole body tomographic model will be constructed by using the procedure developed in this study

  7. THE DISPUTE RESOLUTION MODEL OF VILLAGE HEAD ELECTION THROUGH NON LITIGATION

    Directory of Open Access Journals (Sweden)

    Sri Praptianingsih

    2017-06-01

    Full Text Available Article Number 6 of 2014 clauses 37 verses (5 and (6 provides that the regent in the district must resolve the dispute over the election result of the village head within 30 days. At the district level, the Regional Regulations governing the settlement of village head election disputes and regulations are effective in the dispute profession.However, the laws and regulations at the local / district level have not yet clearly defined the form / format of the outcome of the dispute over the election of the village mayors. The specific purpose of this research is to formulate the model form in the effort to solve the disputes of Village mayors Election by doing syncretism of existing strategy. The Urgency of this research that is (a need to build juridical system in handling dispute of village head election; (b the synchronization of district regulations governing the handling of village head election disputes both vertically and horizontally (c needs a dispute resolution strategy by developing a model of settlement that provides protection of constitutional rights and ensures that government agenda.Research activities in Jember, Bondowoso and Lumajang districts, with a total sample of 150 people. Data collection techniques use Participatory Action Research (PAR and Focus Group Discussion (FGD methods. The Data analysis technique using qualitative analysis.The result of this research is the policy of settlement of disputes of village head election is set forth in juridical instrument at local level, result of settlement stated in peace agreement.This Agreement is then submitted to the Court for the issuance of the Deed of Peace in order to ensure the validity of the legal force for the parties.

  8. 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.

  9. 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

  10. Cyclic loading of thick vessels based on the Prager and Armstrong-Frederick kinematic hardening models

    International Nuclear Information System (INIS)

    Mahbadi, H.; Eslami, M.R.

    2006-01-01

    The aim of this paper is to relate the type of stress category in cyclic loading to ratcheting or shakedown behaviour of the structure. The kinematic hardening theory of plasticity based on the Prager and Armstrong-Frederick models is used to evaluate the cyclic loading behaviour of thick spherical and cylindrical vessels under load and deformation controlled stresses. It is concluded that kinematic hardening based on the Prager model under load and deformation controlled conditions, excluding creep, results in shakedown or reversed plasticity for spherical and cylindrical vessels with the isotropy assumption of the tension/compression curve. Under an anisotropy assumption of the tension/compression curve, this model predicts ratcheting. On the other hand, the Armstrong-Frederick model predicts ratcheting under load controlled cyclic loading and reversed plasticity for deformation controlled stress. The interesting conclusion is that the Armstrong-Frederick model is well capable to predict the experimental data under the assumed type of stresses, wherever experimental data are available

  11. An experimental study of hypervapotron structure in external reactor vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yufeng; Zhang, Ming [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei, E-mail: chenpeipei@snptc.com.cn [State Power Investment Group Corporation, Beijing (China)

    2016-07-15

    Highlights: • Experiments are performed to study the application of hypervapotron in ERVC design. • CHF experiments on two surfaces are conducted under different flow conditions. • Hypervapotron improves CHF performance by 40–60% compared with smooth surface. • Visualization shows fin structure removes vapor mushroom for better liquid supply. - Abstract: In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to

  12. A Semi-analytical model for creep life prediction of butt-welded joints in cylindrical vessels

    International Nuclear Information System (INIS)

    Zarrabi, K.

    2001-01-01

    There have been many investigations on the life assessment of high temperature weldments used in cylindrical pressure vessels, pipes and tubes over the last two decades or so. But to the author's knowledge, currently, there exists no practical, economical and relatively accurate model for creep life assessment of butt-welded joints in cylindrical pressure vessels. This paper describes a semi-analytical and economical model for creep life assessment of butt-welded joints. The first stage of the development of the model is described where the model takes into account the material discontinuities at the welded joint only. The development of the model to include other factors such as geometrical stress concentrations, residual stresses, etc will be reported separately. It has been shown that the proposed model can estimate the redistributions of stresses in the weld and Haz with an error of less than 4%. It has also been shown that the proposed model can conservatively predict the creep life of a butt-welded joint with an error of less than 16%

  13. 78 FR 42733 - Safety Zone; Cleveland Dragon Boat Festival and Head of the Cuyahoga, Cuyahoga River, Cleveland, OH

    Science.gov (United States)

    2013-07-17

    ...-AA00 Safety Zone; Cleveland Dragon Boat Festival and Head of the Cuyahoga, Cuyahoga River, Cleveland... intended to restrict vessels from a portion of the Cuyahoga River during the Dragon Boat Festival and Head... over a decade and the Dragon Boat Festival for the last 7 years. In response to past years' events, the...

  14. Integrated conjugate heat transfer analysis method for in-vessel retention with external reactor vessel cooling - 15477

    International Nuclear Information System (INIS)

    Park, J.W.; Bae, J.H.; Seol, W.C.

    2015-01-01

    An integrated conjugate heat transfer analysis method for the thermal integrity of a reactor vessel under external reactor vessel cooling conditions is developed to resolve light metal layer focusing effect issue. The method calculates steady-state 3-dimensional temperature distribution of a reactor vessel using coupled conjugate heat transfer between in-vessel 3-layered stratified corium (metallic pool, oxide pool and heavy metal) and polar-angle dependent boiling heat transfer at the outer surface of a reactor vessel. The 3-layer corium heat transfer model is utilizing lumped-parameter thermal-resistance circuit method and ex-vessel boiling regimes are parametrically considered. The thermal integrity of a reactor vessel is addressed in terms of un-molten thickness profile. The vessel 3-dimensional heat conduction is validated against a commercial code. It is found that even though the internal heat flux from the metal layer goes far beyond critical heat flux (CHF) the heat flux from the outermost nodes of the vessel may be maintained below CHF due to massive vessel heat diffusion. The heat diffusion throughout the vessel is more pronounced for relatively low heat generation rate in an oxide pool. Parametric calculations are performed considering thermal conditions such as peak heat flux from a light metal layer, heat generation in an oxide pool and external boiling conditions. The major finding is that the most crucial factor for success of in-vessel retention is not the mass of the molten light metal above the oxide pool but the heat generation rate inside the oxide pool and the 3-dimensional vessel heat transfer provides a much larger minimum vessel wall thickness. (authors)

  15. Novel Profiling Model and Side Effects of Helical Scan Silicon Heads

    NARCIS (Netherlands)

    Hozoi, A.; Groenland, J.P.J.; Albertini, J.B.; Lodder, J.C.

    2002-01-01

    Partial erasure of track edges was directly measured from triple-track patterns using a novel model to interpret the output profiles. The model is based on representing the read head as the sum of a reference width, wavelength independent, and two side reading effective widths that are wavelength

  16. The Realistic Versus the Spherical Head Model in EEG Dipole Source Analysis in the Presence of Noise

    National Research Council Canada - National Science Library

    Vanrumste, Bart

    2001-01-01

    .... For 27 electrodes, an EEG epoch of one time sample and spatially white Gaussian noise we found that the importance of the realistic head model over the spherical head model reduces by increasing the noise level.

  17. Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

    2017-04-01

    Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

  18. Analysis of two colliding fractionally damped spherical shells in modelling blunt human head impacts

    Science.gov (United States)

    Rossikhin, Yury A.; Shitikova, Marina V.

    2013-06-01

    The collision of two elastic or viscoelastic spherical shells is investigated as a model for the dynamic response of a human head impacted by another head or by some spherical object. Determination of the impact force that is actually being transmitted to bone will require the model for the shock interaction of the impactor and human head. This model is indended to be used in simulating crash scenarios in frontal impacts, and provide an effective tool to estimate the severity of effect on the human head and to estimate brain injury risks. The model developed here suggests that after the moment of impact quasi-longitudinal and quasi-transverse shock waves are generated, which then propagate along the spherical shells. The solution behind the wave fronts is constructed with the help of the theory of discontinuities. It is assumed that the viscoelastic features of the shells are exhibited only in the contact domain, while the remaining parts retain their elastic properties. In this case, the contact spot is assumed to be a plane disk with constant radius, and the viscoelastic features of the shells are described by the fractional derivative standard linear solid model. In the case under consideration, the governing differential equations are solved analytically by the Laplace transform technique. It is shown that the fractional parameter of the fractional derivative model plays very important role, since its variation allows one to take into account the age-related changes in the mechanical properties of bone.

  19. SAR in human head model due to resonant wireless power transfer system.

    Science.gov (United States)

    Zhang, Chao; Liu, Guoqiang; Li, Yanhong; Song, Xianjin

    2016-04-29

    Efficient mid-range wireless power transfer between transmitter and the receiver has been achieved based on the magnetic resonant coupling method. The influence of electromagnetic field on the human body due to resonant wireless power transfer system (RWPT) should be taken into account during the design process of the system. To analyze the transfer performance of the RWPT system and the change rules of the specific absorption rate (SAR) in the human head model due to the RWPT system. The circuit-field coupling method for a RWPT system with consideration of the displacement current was presented. The relationship between the spiral coil parameters and transfer performance was studied. The SAR in the human head model was calculated under two different exposure conditions. A system with output power higher than 10 W at 0.2 m distance operating at a frequency of approximately 1 MHz was designed. The FEM simulation results show the peak SAR value is below the safety limit which appeared when the human head model is in front of the transmitter. The simulation results agreed well with the experimental results, which verified the validity of the analysis and design.

  20. Physical scale modeling of single free head piles under lateral loading in cohesive soils

    Directory of Open Access Journals (Sweden)

    Edgar Leonardo Salamanca-Medina

    2017-06-01

    Full Text Available This paper presents the results of the small scale modeling of free head wood piles under horizontal loading in cohesive soils, tested in order to compare the results with analytical models proposed by various authors. Characteristic Load (CLM and P-Y Curves methods were used for the prediction of lateral deflections at the head of the piles and the method proposed by Broms for estimating the ultimate lateral load. These predictions were compared with the results of the physical modeling, obtaining a good approximation between them.

  1. Development of New, Low-Head Hydropower Turbine - Modeling & Laboratory Test DE-EE0005426

    Energy Technology Data Exchange (ETDEWEB)

    Krouse, Wayne [Hydro Green Energy, Westmont, IL (United States)

    2014-12-05

    Hydro Green Energy, LLC (HGE) will complete the design, fabrication and laboratory testing of a scaled, vertically stackable, low-head hydropower turbine called the Modular Bulb Turbine (MBT). HGE will also complete a summary report that includes the laboratory testing results and analysis of the tests. Project Goals: Design, model and test modular bulb turbine for installation in numerous HGE low-head hydropower projects at non-powered USACE dams. Project Results: The sub-scale prototype was tested successfully at a leading US hydraulic laboratory. Laboratory data results agreed well with predicted results from numerical modeling.

  2. Osteonecrosis of femoral head: Treatment by core decompression and vascular pedicle grafting

    Directory of Open Access Journals (Sweden)

    Babhulkar Sudhir

    2009-01-01

    Full Text Available Background: Femoral head-preserving core decompression and bone grafting have shown excellent result in preventing collapse. The use of vascularized grafts have shown better clinical results. The vascular pedicle bone graft is an easy to perform operation and does not require special equipment. We analyzed and report a series of patients of osteonecrosis of femoral head treated by core decompression and vascular pedicle grafting of part of iliac crest based on deep circumflex iliac vessels. Materials and Methods: The article comprises of the retrospective study of 31 patients of osteonecrosis of femoral head in stage II and III treated with core decompression and vascular pedicle grafting by using part of iliac crest with deep circumflex iliac vessels from January 1990 to December 2005. The young patients with a mean age 32 years (18-52 years with a minimum follow-up of five years were included for analysis. Sixteen patients had osteonecrosis following alcohol abuse, 12 patients following corticosteroid consumption, 3 patients had idiopathic osteonecrosis. Nine patients were stage IIB, and 22 patients were stage IIIC according to ARCO′s system. The core decompression and vascular pedicle grafting was performed by anterior approach by using part of iliac crest with deep circumflex iliac vessels. Results: Digital subtraction arteriography performed in 9 patients at the end of 12 weeks showed the patency of deep circumflex artery in all cases, and bone scan performed in 6 other patients showed high uptake in the grafted area of the femoral head proving the efficacy of the operative procedure. Out of 31 patients, only one patient progressed to collapse and total joint replacement was advised. At the final follow up period of 5-8 years, Harris Hip Score improved mean ± SD of 28.2 ± 6.4 ( p < 0.05. Forty-eight percent of patients had an improvement in Harris Hip Score of more that 28 points. Conclusion: The core decompression and vascular pedicle

  3. Vascular endothelial growth factor/bone morphogenetic protein-2 bone marrow combined modification of the mesenchymal stem cells to repair the avascular necrosis of the femoral head

    Science.gov (United States)

    Ma, Xiao-Wei; Cui, Da-Ping; Zhao, De-Wei

    2015-01-01

    Vascular endothelial cell growth factor (VEGF) combined with bone morphogenetic protein (BMP) was used to repair avascular necrosis of the femoral head, which can maintain the osteogenic phenotype of seed cells, and effectively secrete VEGF and BMP-2, and effectively promote blood vessel regeneration and contribute to formation and revascularization of tissue engineered bone tissues. To observe the therapeutic effect on the treatment of avascular necrosis of the femoral head by using bone marrow mesenchymal stem cells (BMSCs) modified by VEGF-165 and BMP-2 in vitro. The models were avascular necrosis of femoral head of rabbits on right leg. There groups were single core decompression group, core decompression + BMSCs group, core decompression + VEGF-165/BMP-2 transfect BMSCs group. Necrotic bone was cleared out under arthroscope. Arthroscopic observation demonstrated that necrotic bone was cleared out in each group, and fresh blood flowed out. Histomorphology determination showed that blood vessel number and new bone area in the repair region were significantly greater at various time points following transplantation in the core decompression + VEGF-165/BMP-2 transfect BMSCs group compared with single core decompression group and core decompression + BMSCs group (P < 0.05). These suggested that VEGF-165/BMP-2 gene transfection strengthened osteogenic effects of BMSCs, elevated number and quality of new bones and accelerated the repair of osteonecrosis of the femoral head. PMID:26629044

  4. An experimental study of assessment of weld quality on fatigue reliability analysis of a nuclear pressure vessel

    International Nuclear Information System (INIS)

    Dai Shuhe

    1993-01-01

    The steam generator in PWR primary coolant system China of Qinshan Nuclear Power Plant is a crucial unit belonging to the category of nuclear pressure vessel. The purpose of this research work is to make an examination of the weld quality of the steam generator under fatigue loading and to assess its reliability by using the experimental results of fatigue test of material of nuclear pressure vessel S-271 (Chinese Standard) and of qualified tests of welded seams of a simulated prototype of bottom closure head of the steam generator. A guarantee of weld quality is proposed as a subsequent verification for China National Nuclear Safety Supervision Bureau. The results of reliability analysis reported in this work can be taken as a supplementary material of Probabilistic Safety Assessment (PSA) of Qinshan Nuclear Power Plant. According to the requirement of Provision II-1500 cyclic testing, ASME Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Power Plant Components, a simulated prototype of the bottom closure head of the steam generator was made for qualified tests. To find the quantified results of reliability assessment by using the testing data, two proposals are presented

  5. A Measure of Similarity Between Trajectories of Vessels

    Directory of Open Access Journals (Sweden)

    Le QI

    2016-03-01

    Full Text Available The measurement of similarity between trajectories of vessels is one of the kernel problems that must be addressed to promote the development of maritime intelligent traffic system (ITS. In this study, a new model of trajectory similarity measurement was established to improve the data processing efficiency in dynamic application and to reflect actual sailing behaviors of vessels. In this model, a feature point detection algorithm was proposed to extract feature points, reduce data storage space and save computational resources. A new synthesized distance algorithm was also created to measure the similarity between trajectories by using the extracted feature points. An experiment was conducted to measure the similarity between the real trajectories of vessels. The growth of these trajectories required measurements to be conducted under different voyages. The results show that the similarity measurement between the vessel trajectories is efficient and correct. Comparison of the synthesized distance with the sailing behaviors of vessels proves that results are consistent with actual situations. The experiment results demonstrate the promising application of the proposed model in studying vessel traffic and in supplying reliable data for the development of maritime ITS.

  6. A three-temperature model of selective photothermolysis for laser treatment of port wine stain containing large malformed blood vessels

    International Nuclear Information System (INIS)

    Li, D.; Wang, G.X.; He, Y.L.; Wu, W.J.; Chen, B.

    2014-01-01

    As congenital vascular malformations, port wine stain (PWS) is composed of ectatic venular capillary blood vessels buried within healthy dermis. In clinic, pulsed dye laser (PDL) in visible band (e.g. 585 nm) together with cryogen spray cooling (CSC) have become the golden standard for treatment of PWS. However, due to the limited energy deposition of the PDL in blood, large blood vessels are likely to survive from the laser irradiation. As a result, complete clearance of the lesions is rarely achieved. Assuming the local thermal non-equilibrium in skin tissue during the laser surgery, a three-temperature model is proposed to treat the PWS tissue as a porous media composed of a non-absorbing dermal matrix buried with the blood as well as the large malformed blood vessels. Three energy equations are constructed and solved coupling for the temperature of the blood in average-sized PWS vessels, non-absorbing dermal tissues and large malformed blood vessels, respectively. Subsequently, the thermal responses of human skin to visible (585 nm) and near-infrared (1064 nm) laser irradiations with various pulse durations in conjunction with cryogen spray cooling are investigated by the new model, and Arrhenius integral is used to analyze the thermal damage. The simulations show that the short pulse duration of 1.5 ms results in a higher selective heating of blood over epidermis, which will lead to a desired clinic outcome than the longer pulse duration. Due to a much deeper light penetration depth, laser irradiation with 1064 nm in wavelength is superior to that with 585 nm in treating patients with cutaneous hyper-vascular malformation. Complete coagulations are predicted in large-sized and deeply extending blood vessels by 1064 nm laser. - Highlights: •A three-temperature model is proposed for the laser treatment of port wine stain (PWS). •Average sized and large malformed blood vessels in porous medium (tissue) are considered. •Thermal responses of PWS to

  7. Constitutive models for concrete and finite element analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Smith, P.D.; Anderson, C.A.

    1977-01-01

    Two constitutive models for concrete are discussed. For short-term loads, the orthotropic variable modulus model is described, and for long-term loads a viscoelastic model utilizing a Dirichlet series approximation for the creep compliance function is summarized. The orthotropic variable modulus model is demonstrated in an analysis of a PCRV head with penetrations. The viscoelastic model is illustrated with a simulation of a prestressed concrete cylinder subject to non-uniform temperatures

  8. In-vessel coolability and retention of a core melt

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Liu, C.; Additon, S.

    1997-01-01

    The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. The technical treatment in this assessment includes: (a) new data on energy flow from either volumetrically heated pools or non-heated layers on top, boiling and critical heat flux in inverted, curved geometries, emissivity of molten (superheated) samples of steel, and chemical reactivity proof tests, (b) a simple but accurate mathematical formulation that allows prediction of thermal loads by means of convenient hand calculations, (c) a detailed model programmed on the computer to sample input parameters over the uncertainty ranges, and to produce probability distributions of thermal loads and margins for departure from nucleate boiling at each angular position on the lower head, and (d) detailed structural evaluations that demonstrate that departure from nucleate boiling is a necessary and sufficient criterion for failure. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is open-quotes physically unreasonable.close quotes Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings

  9. Effects of blood-activating and stasis-removing drugs combined with VEGF gene transfer on angiogenesis in ischemic necrosis of the femoral head.

    Science.gov (United States)

    Li, Jun-Hui; Wu, Ya-Ling; Ye, Jian-Hong; Ning, Ya-Gong; Yu, Hai-Ying; Peng, Zhong-Jie; Luan, Xiao-Wen

    2009-09-01

    To observe the promoting effects of blood-activating and stasis-removing Chinese drugs combined with vascular endothelial growth factor (VEGF) gene transfer on angiogenesis in ischemic necrosis of the femoral head. Forty Japanese giant-ear rabbits were randomly divided into a control group, a model group, a Chinese drug group, a gene group, and a combined group. After 8 weeks of treatment, the rate of VEGF positive cell expression in the synovium of the femoral head was measured using the immunohistochemical method, and the number of blood vessels in the femoral head was measured by digital subtraction angiography. The rate of VEGF positive cell expression in the model group was significantly lower than that in the Chinese drug group (P 0.05). Either the blood-activating and stasis-removing Chinese drugs or VEGF gene transfer can promote the angiogenesis and building of collateral circulation for femoral head ischemic necrosis, and the combined therapy with Chinese drugs or VEGF gene transfer may show a better therapeutic effect. The present study provides an experimental basis for clinical application of the combined therapy with the blood-activating and stasis-removing Chinese drugs and VEGF gene transfer.

  10. Probabilistic atlas based labeling of the cerebral vessel tree

    Science.gov (United States)

    Van de Giessen, Martijn; Janssen, Jasper P.; Brouwer, Patrick A.; Reiber, Johan H. C.; Lelieveldt, Boudewijn P. F.; Dijkstra, Jouke

    2015-03-01

    Preoperative imaging of the cerebral vessel tree is essential for planning therapy on intracranial stenoses and aneurysms. Usually, a magnetic resonance angiography (MRA) or computed tomography angiography (CTA) is acquired from which the cerebral vessel tree is segmented. Accurate analysis is helped by the labeling of the cerebral vessels, but labeling is non-trivial due to anatomical topological variability and missing branches due to acquisition issues. In recent literature, labeling the cerebral vasculature around the Circle of Willis has mainly been approached as a graph-based problem. The most successful method, however, requires the definition of all possible permutations of missing vessels, which limits application to subsets of the tree and ignores spatial information about the vessel locations. This research aims to perform labeling using probabilistic atlases that model spatial vessel and label likelihoods. A cerebral vessel tree is aligned to a probabilistic atlas and subsequently each vessel is labeled by computing the maximum label likelihood per segment from label-specific atlases. The proposed method was validated on 25 segmented cerebral vessel trees. Labeling accuracies were close to 100% for large vessels, but dropped to 50-60% for small vessels that were only present in less than 50% of the set. With this work we showed that using solely spatial information of the vessel labels, vessel segments from stable vessels (>50% presence) were reliably classified. This spatial information will form the basis for a future labeling strategy with a very loose topological model.

  11. Experimental analysis of a nuclear reactor prestressed concrete pressure vessels model

    International Nuclear Information System (INIS)

    Vallin, C.

    1980-01-01

    A comprehensible analysis was made of the performance of each set of sensors used to measure the strain and displacement of a 1/20 scale Prestressed Concrete Pressure Vessel (PCPV) model tested at the Instituto de Pesquisas Energeticas e Nucleares (IPEN). Among the three Kinds of sensors used (strain gage, displacement transducers and load cells) the displacement transducers showed the best behavior. The displacemente transducers data was statistically analysed and a linear behavior of the model was observed during the first pressurizations tests. By means of a linear statistical correlation between experimental and expected theoretical data it was found that the model looses the linearity at a pressure between 110-125 atm. (Author) [pt

  12. Revascularization of femoral head ischemic necrosis with vascularized bone graft: A CT scan experimental study

    International Nuclear Information System (INIS)

    Gonzalez del Pino, J.; Knapp, K.; Gomez Castresana, F.; Benito, M.

    1990-01-01

    An ischemic necrosis of the femoral head was induced in 15 mongrel adult dogs using the technique described by Gartsman et al. Five weeks later, a free vascularized rib graft was transferred into the previously induced ischemic femoral head. High resolution computed tomographic scanning was used to evaluate revascularization 4, 8 and 12 weeks after grafting. The femoral head exhibited new vessel formation throughout the study. Arterial terminal branches arising from the rib graft medullary and periosteal circulations extended beyond the rib graft, entered the head, and reached the subchondral plate. Even where the rib graft did not replenish the central core of the head, there was vascular supply from the grafted bone's vascular tree. These results suggest that a free vascularized bone graft is able to revascularize an experimentally induced ischemic femoral head necrosis. (orig.)

  13. Thermal mathematical modeling of a multicell common pressure vessel nickel-hydrogen battery

    Science.gov (United States)

    Kim, Junbom; Nguyen, T. V.; White, R. E.

    1992-01-01

    A two-dimensional and time-dependent thermal model of a multicell common pressure vessel (CPV) nickel-hydrogen battery was developed. A finite element solver called PDE/Protran was used to solve this model. The model was used to investigate the effects of various design parameters on the temperature profile within the cell. The results were used to help find a design that will yield an acceptable temperature gradient inside a multicell CPV nickel-hydrogen battery. Steady-state and unsteady-state cases with a constant heat generation rate and a time-dependent heat generation rate were solved.

  14. A mouse model of weight-drop closed head injury: emphasis on cognitive and neurological deficiency

    Directory of Open Access Journals (Sweden)

    Igor Khalin

    2016-01-01

    Full Text Available Traumatic brain injury (TBI is a leading cause of death and disability in individuals worldwide. Producing a clinically relevant TBI model in small-sized animals remains fairly challenging. For good screening of potential therapeutics, which are effective in the treatment of TBI, animal models of TBI should be established and standardized. In this study, we established mouse models of closed head injury using the Shohami weight-drop method with some modifications concerning cognitive deficiency assessment and provided a detailed description of the severe TBI animal model. We found that 250 g falling weight from 2 cm height produced severe closed head injury in C57BL/6 male mice. Cognitive disorders in mice with severe closed head injury could be detected using passive avoidance test on day 7 after injury. Findings from this study indicate that weight-drop injury animal models are suitable for further screening of brain neuroprotectants and potentially are similar to those seen in human TBI.

  15. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, Pritam [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Biner, Suleyman Bulent [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Spencer, Benjamin Whiting [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-07-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  16. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    International Nuclear Information System (INIS)

    Chakraborty, Pritam; Biner, Suleyman Bulent; Zhang, Yongfeng; Spencer, Benjamin Whiting

    2015-01-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  17. The modeling of core melting and in-vessel corium relocation in the APRIL code

    Energy Technology Data Exchange (ETDEWEB)

    Kim. S.W.; Podowski, M.Z.; Lahey, R.T. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others

    1995-09-01

    This paper is concerned with the modeling of severe accident phenomena in boiling water reactors (BWR). New models of core melting and in-vessel corium debris relocation are presented, developed for implementation in the APRIL computer code. The results of model testing and validations are given, including comparisons against available experimental data and parametric/sensitivity studies. Also, the application of these models, as parts of the APRIL code, is presented to simulate accident progression in a typical BWR reactor.

  18. Development of FB-MultiPier dynamic vessel-collision analysis models, phase 2.

    Science.gov (United States)

    2014-07-01

    Massive waterway vessels such as barges regularly transit navigable waterways in the U.S. During passages that fall within : the vicinity of bridge structures, vessels may (under extreme circumstances) deviate from the intended vessel transit path. A...

  19. SCDAP/RELAP5 modeling of movement of melted material through porous debris in lower head

    International Nuclear Information System (INIS)

    Siefken, L. J.; Harvego, E. A.

    2000-01-01

    A model is described for the movement of melted metallic material through a ceramic porous debris bed. The model is designed for the analysis of severe accidents in LWRs, wherein melted core plate material may slump onto the top of a porous bed of relocated core material supported by the lower head. The permeation of the melted core plate material into the porous debris bed influences the heatup of the debris bed and the heatup of the lower head supporting the debris. A model for mass transport of melted metallic material is applied that includes terms for viscosity and turbulence but neglects inertial and capillary terms because of their small value relative to gravity and viscous terms in the momentum equation. The relative permeability and passability of the porous debris are calculated as functions of debris porosity, particle size, and effective saturation. An iterative numerical solution is used to solve the set of nonlinear equations for mass transport. The effective thermal conductivity of the debris is calculated as a function of porosity, particle size, and saturation. The model integrates the equations for mass transport with a model for the two-dimensional conduction of heat through porous debris. The integrated model has been implemented into the SCDAP/RELAP5 code for the analysis of the integrity of LWR lower heads during severe accidents. The results of the model indicate that melted core plate material may permeate to near the bottom of a 1m deep hot porous debris bed supported by the lower head. The presence of the relocated core plate material was calculated to cause a 12% increase in the heat flux on the external surface of the lower head

  20. Effort dynamics in a fisheries bioeconomic model: A vessel level approach through Game Theory

    Directory of Open Access Journals (Sweden)

    Gorka Merino

    2007-09-01

    Full Text Available Red shrimp, Aristeus antennatus (Risso, 1816 is one of the most important resources for the bottom-trawl fleets in the northwestern Mediterranean, in terms of both landings and economic value. A simple bioeconomic model introducing Game Theory for the prediction of effort dynamics at vessel level is proposed. The game is performed by the twelve vessels exploiting red shrimp in Blanes. Within the game, two solutions are performed: non-cooperation and cooperation. The first is proposed as a realistic method for the prediction of individual effort strategies and the second is used to illustrate the potential profitability of the analysed fishery. The effort strategy for each vessel is the number of fishing days per year and their objective is profit maximisation, individual profits for the non-cooperative solution and total profits for the cooperative one. In the present analysis, strategic conflicts arise from the differences between vessels in technical efficiency (catchability coefficient and economic efficiency (defined here. The ten-year and 1000-iteration stochastic simulations performed for the two effort solutions show that the best strategy from both an economic and a conservationist perspective is homogeneous effort cooperation. However, the results under non-cooperation are more similar to the observed data on effort strategies and landings.

  1. A study of reactor vessel integrity assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Hoon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Kim, Jong Kyung; Shin, Chang Ho; Seo, Bo Kyun [Hanyang Univ., Seoul (Korea, Republic of)

    1999-02-15

    The fast neutron fluence at the Reactor Pressure Vessel(RPV) of KNGR designed for 60 years lifetime was calculated by full-scope Monte Carlo simulation for reactor vessel integrity assessment. KNGR core geometry was modeled on a three-dimensional representation of the one-sixteenth of the reactor in-vessel component. Each fuel assemblies were modeled explicitly, and each fuel pins were axially divided into 5 segments. The maximum flux of 4.3 x 10{sup 10} neutrons/cm{sup 2}. sec at the RPV was obtained by tallying neutrons crossing the beltline of inner surface of the RPV.

  2. Gas-liquid flow filed in agitated vessels

    International Nuclear Information System (INIS)

    Hormazi, F.; Alaie, M.; Dabir, B.; Ashjaie, M.

    2001-01-01

    Agitated vessels in form of sti reed tank reactors and mixed ferment ors are being used in large numbers of industry. It is more important to develop good, and theoretically sound models for scaling up and design of agitated vessels. In this article, two phase flow (gas-liquid) in a agitated vessel has been investigated numerically. A two-dimensional computational fluid dynamics model, is used to predict the gas-liquid flow. The effects of gas phase, varying gas flow rates and variation of bubbles shape on flow filed of liquid phase are investigated. The numerical results are verified against the experimental data

  3. Development of an in silico stochastic 4D model of tumor growth with angiogenesis.

    Science.gov (United States)

    Forster, Jake C; Douglass, Michael J J; Harriss-Phillips, Wendy M; Bezak, Eva

    2017-04-01

    A stochastic computer model of tumour growth with spatial and temporal components that includes tumour angiogenesis was developed. In the current work it was used to simulate head and neck tumour growth. The model also provides the foundation for a 4D cellular radiotherapy simulation tool. The model, developed in Matlab, contains cell positions randomised in 3D space without overlap. Blood vessels are represented by strings of blood vessel units which branch outwards to achieve the desired tumour relative vascular volume. Hypoxic cells have an increased cell cycle time and become quiescent at oxygen tensions less than 1 mmHg. Necrotic cells are resorbed. A hierarchy of stem cells, transit cells and differentiated cells is considered along with differentiated cell loss. Model parameters include the relative vascular volume (2-10%), blood oxygenation (20-100 mmHg), distance from vessels to the onset of necrosis (80-300 μm) and probability for stem cells to undergo symmetric division (2%). Simulations were performed to observe the effects of hypoxia on tumour growth rate for head and neck cancers. Simulations were run on a supercomputer with eligible parts running in parallel on 12 cores. Using biologically plausible model parameters for head and neck cancers, the tumour volume doubling time varied from 45 ± 5 days (n = 3) for well oxygenated tumours to 87 ± 5 days (n = 3) for severely hypoxic tumours. The main achievements of the current model were randomised cell positions and the connected vasculature structure between the cells. These developments will also be beneficial when irradiating the simulated tumours using Monte Carlo track structure methods. © 2017 American Association of Physicists in Medicine.

  4. NEAR REAL-TIME AUTOMATIC MARINE VESSEL DETECTION ON OPTICAL SATELLITE IMAGES

    Directory of Open Access Journals (Sweden)

    G. Máttyus

    2013-05-01

    Full Text Available Vessel monitoring and surveillance is important for maritime safety and security, environment protection and border control. Ship monitoring systems based on Synthetic-aperture Radar (SAR satellite images are operational. On SAR images the ships made of metal with sharp edges appear as bright dots and edges, therefore they can be well distinguished from the water. Since the radar is independent from the sun light and can acquire images also by cloudy weather and rain, it provides a reliable service. Vessel detection from spaceborne optical images (VDSOI can extend the SAR based systems by providing more frequent revisit times and overcoming some drawbacks of the SAR images (e.g. lower spatial resolution, difficult human interpretation. Optical satellite images (OSI can have a higher spatial resolution thus enabling the detection of smaller vessels and enhancing the vessel type classification. The human interpretation of an optical image is also easier than as of SAR image. In this paper I present a rapid automatic vessel detection method which uses pattern recognition methods, originally developed in the computer vision field. In the first step I train a binary classifier from image samples of vessels and background. The classifier uses simple features which can be calculated very fast. For the detection the classifier is slided along the image in various directions and scales. The detector has a cascade structure which rejects most of the background in the early stages which leads to faster execution. The detections are grouped together to avoid multiple detections. Finally the position, size(i.e. length and width and heading of the vessels is extracted from the contours of the vessel. The presented method is parallelized, thus it runs fast (in minutes for 16000 × 16000 pixels image on a multicore computer, enabling near real-time applications, e.g. one hour from image acquisition to end user.

  5. Near Real-Time Automatic Marine Vessel Detection on Optical Satellite Images

    Science.gov (United States)

    Máttyus, G.

    2013-05-01

    Vessel monitoring and surveillance is important for maritime safety and security, environment protection and border control. Ship monitoring systems based on Synthetic-aperture Radar (SAR) satellite images are operational. On SAR images the ships made of metal with sharp edges appear as bright dots and edges, therefore they can be well distinguished from the water. Since the radar is independent from the sun light and can acquire images also by cloudy weather and rain, it provides a reliable service. Vessel detection from spaceborne optical images (VDSOI) can extend the SAR based systems by providing more frequent revisit times and overcoming some drawbacks of the SAR images (e.g. lower spatial resolution, difficult human interpretation). Optical satellite images (OSI) can have a higher spatial resolution thus enabling the detection of smaller vessels and enhancing the vessel type classification. The human interpretation of an optical image is also easier than as of SAR image. In this paper I present a rapid automatic vessel detection method which uses pattern recognition methods, originally developed in the computer vision field. In the first step I train a binary classifier from image samples of vessels and background. The classifier uses simple features which can be calculated very fast. For the detection the classifier is slided along the image in various directions and scales. The detector has a cascade structure which rejects most of the background in the early stages which leads to faster execution. The detections are grouped together to avoid multiple detections. Finally the position, size(i.e. length and width) and heading of the vessels is extracted from the contours of the vessel. The presented method is parallelized, thus it runs fast (in minutes for 16000 × 16000 pixels image) on a multicore computer, enabling near real-time applications, e.g. one hour from image acquisition to end user.

  6. The autopsy-correlation of computed tomography in acute severe head injuries

    International Nuclear Information System (INIS)

    Tomita, Shin; Kim, Hong; Mikabe, Toshio; Karasawa, Hideharu; Watanabe, Saburo

    1981-01-01

    We discuss the importance of Contrast-Enhanced CT (C.E.CT) in establishing the variety of the intracranial pathological process in acute severe head injuries. During a two-and-a-half-year period (June, 1977 - December, 1979) thirty-three patients with acute severe head injuries were autopsied, all of whom had been scanned on admission. Among them, 14 patients had undergone both plain CT and C.E.CT on admission. Brain slices were examined macroscopically in three categories; brain contusion, subarachnoid hemorrhage, and intracerebral hemorrhage. Each category was then compared retrospectively with the plain CT and C.E.CT findings. C.E.CT was found to correspond much better to the autopsy finding than plain CT in the following three points: (1) C.E.CT clearly enhances the contusion areas and reveals occult contusion areas. (2) C.E.CT enhances the areas corresponding to the subarachnoid space due to the breakdown of brain-surface blood vessels. (3) C.E.CT reveals the enlargement and formation of the intracerebral hematoma by the extravasation of the intravenous contrast material from injured arterial vessels. (author)

  7. Femoral head vitality after intracapsular hip fracture

    International Nuclear Information System (INIS)

    Stroemqvist, B.

    1983-01-01

    Femoral head vitality before, during and at various intervals from the operation was determined by tetracycline labeling and/or 99 sp (m)Tc-MDP scintimetry. In a three-year follow-up, healing prognosis could be determined by scintimetry 3 weeks from operation; deficient femoral head vitality predicting healing complications and retained vitality predicting uncomplicated healing. A comparison between pre- and postoperative scintimetry indicated that further impairment of the femoral head vitality could be caused by the operative procedure, and as tetracycline labeling prior to and after fracture reduction in 370 fractures proved equivalent, it was concluded that the procedure of osteosynthesis probably was responsible for capsular vessel injury, using a four-flanged nail. The four-flanged nail was compared with a low-traumatic method of osteosynthesis, two hook-pins, in a prospective randomized 14 month study, and the postoperative femoral head vitality was significantly better in the hook-pin group. This was also clearly demonstrated in a one-year follow-up for the fractures included in the study. Parallel to these investigations, the reliability of the methods of vitality determination was found satisfactory in methodologic studies. For clinical purpose, primary atraumatic osteosynthesis, postoperative prognostic scintimetry and early secondary arthroplasty when indicated, was concluded to be the appropriate approach to femoral neck fracture treatment. (Author)

  8. Case of Small Vessel Disease Associated with COL4A1 Mutations following Trauma

    Directory of Open Access Journals (Sweden)

    Joao McONeil Plancher

    2015-06-01

    Full Text Available With this case report, we would like to heighten the awareness of clinicians about COL4A1 as a single-gene disorder causing cerebral small vessel disease and describe a previously unreported pathogenic missense substitution in COL4A1 (p.Gly990Val and a new clinical presentation. We identified a heterozygous putatively pathogenic mutation of COL4A1 in a 50-year-old female with a history of congenital cataracts and glaucoma who presented with multiple diffusion-positive infarcts and areas of contrast enhancement following mild head trauma. We believe that this presentation of multiple areas of acute brain and vascular injury in the setting of mild head trauma is a new manifestation of this genetic disorder. Imaging findings of multiple acute infarcts and regions of contrast enhancement with associated asymptomatic old deep microhemorrhages and leukomalacia in adults after head trauma should raise a high suspicion for a COL4A1 genetic disorder. Radiographic patterns of significant leukoaraiosis and deep microhemorrhages can also be seen in patients with long-standing vasculopathy associated with hypertension, which our patient lacked. Our findings demonstrate the utility of genetic screening for COL4A1 mutations in young patients who have small vessel vasculopathy on brain imaging but who do not have significant cardiovascular risk factors.

  9. Model of large scale man-machine systems with an application to vessel traffic control

    NARCIS (Netherlands)

    Wewerinke, P.H.; van der Ent, W.I.; ten Hove, D.

    1989-01-01

    Mathematical models are discussed to deal with complex large-scale man-machine systems such as vessel (air, road) traffic and process control systems. Only interrelationships between subsystems are assumed. Each subsystem is controlled by a corresponding human operator (HO). Because of the

  10. Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel. Pt. 2

    International Nuclear Information System (INIS)

    Rose, P.W.

    1987-12-01

    In previous experiments, done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel, the instrument tubes support structure built into the vault was not included. It consists of a number of grids made up of fairly massive steel girders. These have now been added to the model and experiments performed using water to simulate molten core debris assumed to have fallen on to the vault floor and high-pressure air to simulate the discharge of steam or gas from the assumed breach at the bottom of the pressure vessel. The results show that the tubes support structure considerably reduces the carry-over of liquid via the vault access shafts. (author)

  11. Interventional neuroradiology of the head and neck.

    Science.gov (United States)

    Turowski, Bernd; Zanella, Friedhelm E

    2003-08-01

    Vascular interventions are important and helpful for treatment of various pathologies of the head and neck. Interventional neuroradiology of the head and neck includes image-guided biopsies, vessel occlusion, and local chemotherapy. Knowledge of anatomy, functional relationships between intra- and extracranial vessels, and pathology are the basis for therapeutic success. The interventional neuroradiologist is responsible for appropriate selection of patients based on clinical information, indications, and risk assessment. Neuroradiologic imaging, especially CT and MR imaging, and appropriate analysis of angiographic findings help ensure indication for treatment and plan an intervention. Technical equipment, including an angiographic unit, catheters, needles, embolizing materials, and so forth, are important. Knowledge of hemodynamics is relevant to avoid complications and to find the optimal technique for solving the clinical problem. Indications for image-guided biopsies are preverterbal fluid-collections, spinal and paraspinal inflammations and abscesses, deep cervical malignancies, vertebral body, and skull base tumors. Special care should be taken to preserve critical structures in this region, including spinal nerve roots, cervical plexus, main peripheral nerves, and vessels. Indications for vessel occlusion are emergency situations to stop bleeding in vascular lesions (traumatic, malformation, or tumors) by reduction of pressure, preoperative reduction of blood flow to minimize the surgical risk, palliative occlusion of feeding vessels to produce tumor necrosis, or potential curative (or presurgical) occlusion of vascular malformations. Pressure reduction to support normal coagulation, such as epistaxis, in hereditary hemorrhagic telangiectasia can be achieved by proximal vessel occlusion with large particles or platinum coils. Prevention of intraoperative bleeding requires occlusion of the microvascular bed with small particles. Examples of these

  12. HYDROïD humanoid robot head with perception and emotion capabilities :Modeling, Design and Experimental Results

    Directory of Open Access Journals (Sweden)

    Samer eAlfayad

    2016-04-01

    Full Text Available In the framework of the HYDROïD humanoid robot project, this paper describes the modeling and design of an electrically actuated head mechanism. Perception and emotion capabilities are considered in the design process. Since HYDROïD humanoid robot is hydraulically actuated, the choice of electrical actuation for the head mechanism addressed in this paper is justified. Considering perception and emotion capabilities leads to a total number of 15 degrees of freedom for the head mechanism which are split on four main sub-mechanisms: the neck, the mouth, the eyes and the eyebrows. Biological data and kinematics performances of human head are taken as inputs of the design process. A new solution of uncoupled eyes is developed to possibly address the master-slave process that links the human eyes as well as vergence capabilities. Modeling each sub-system is carried out in order to get equations of motion, their frequency responses and their transfer functions. The neck pitch rotation is given as a study example. Then, the head mechanism performances are presented through a comparison between model and experimental results validating the hardware capabilities. Finally, the head mechanism is integrated on the HYDROïD upper-body. An object tracking experiment coupled with emotional expressions is carried out to validate the synchronization of the eye rotations with the body motions.

  13. Multivariable modeling of pressure vessel and piping J-R data

    International Nuclear Information System (INIS)

    Eason, E.D.; Wright, J.E.; Nelson, E.E.

    1991-05-01

    Multivariable models were developed for predicting J-R curves from available data, such as material chemistry, radiation exposure, temperature, and Charpy V-notch energy. The present work involved collection of public test data, application of advanced pattern recognition tools, and calibration of improved multivariable models. Separate models were fitted for different material groups, including RPV welds, Linde 80 welds, RPV base metals, piping welds, piping base metals, and the combined database. Three different types of models were developed, involving different combinations of variables that might be available for applications: a Charpy model, a preirradiation Charpy model, and a copper-fluence model. In general, the best results were obtained with the preirradiation Charpy model. The copper-fluence model is recommended only if Charpy data are unavailable, and then only for Linde 80 welds. Relatively good fits were obtained, capable of predicting the values of J for pressure vessel steels to with a standard deviation of 13--18% over the range of test data. The models were qualified for predictive purposes by demonstrating their ability to predict validation data not used for fitting. 20 refs., 45 figs., 16 tabs

  14. Modeling the Role of the Glymphatic Pathway and Cerebral Blood Vessel Properties in Alzheimer's Disease Pathogenesis.

    Science.gov (United States)

    Kyrtsos, Christina Rose; Baras, John S

    2015-01-01

    Alzheimer's disease (AD) is the most common cause of dementia in the elderly, affecting over 10% population over the age of 65 years. Clinically, AD is described by the symptom set of short term memory loss and cognitive decline, changes in mentation and behavior, and eventually long-term memory deficit as the disease progresses. On imaging studies, significant atrophy with subsequent increase in ventricular volume have been observed. Pathology on post-mortem brain specimens demonstrates the classic findings of increased beta amyloid (Aβ) deposition and the presence of neurofibrillary tangles (NFTs) within affected neurons. Neuroinflammation, dysregulation of blood-brain barrier transport and clearance, deposition of Aβ in cerebral blood vessels, vascular risk factors such as atherosclerosis and diabetes, and the presence of the apolipoprotein E4 allele have all been identified as playing possible roles in AD pathogenesis. Recent research has demonstrated the importance of the glymphatic system in the clearance of Aβ from the brain via the perivascular space surrounding cerebral blood vessels. Given the variety of hypotheses that have been proposed for AD pathogenesis, an interconnected, multilayer model offers a unique opportunity to combine these ideas into a single unifying model. Results of this model demonstrate the importance of vessel stiffness and heart rate in maintaining adequate clearance of Aβ from the brain.

  15. Emergency venting of pressure vessels

    International Nuclear Information System (INIS)

    Steinkamp, H.

    1995-01-01

    With the numerical codes developed for safety analysis the venting of steam vessel can be simulated. ATHLET especially is able to predict the void fraction depending on the vessel height. Although these codes contain a one-dimensional model they allow the description of complex geometries due to the detailed nodalization of the considered apparatus. In chemical reactors, however, the venting process is not only influenced by the flashing behaviour but additionally by the running chemical reaction in the vessel. Therefore the codes used for modelling have to consider the kinetics of the chemical reaction. Further multi-component systems and dissolving processes have to be regarded. In order to preduct the fluid- and thermodynamic process it could be helpful to use 3-dimensional codes in combination with the one-dimensional codes as used in nuclear industry to get a more detailed describtion of the running processes. (orig./HP)

  16. Advanced toroidal facility vaccuum vessel stress analyses

    International Nuclear Information System (INIS)

    Hammonds, C.J.; Mayhall, J.A.

    1987-01-01

    The complex geometry of the Advance Toroidal Facility (ATF) vacuum vessel required special analysis techniques in investigating the structural behavior of the design. The response of a large-scale finite element model was found for transportation and operational loading. Several computer codes and systems, including the National Magnetic Fusion Energy Computer Center Cray machines, were implemented in accomplishing these analyses. The work combined complex methods that taxed the limits of both the codes and the computer systems involved. Using MSC/NASTRAN cyclic-symmetry solutions permitted using only 1/12 of the vessel geometry to mathematically analyze the entire vessel. This allowed the greater detail and accuracy demanded by the complex geometry of the vessel. Critical buckling-pressure analyses were performed with the same model. The development, results, and problems encountered in performing these analyses are described. 5 refs., 3 figs

  17. Robot-Assisted Free Flap in Head and Neck Reconstruction

    Directory of Open Access Journals (Sweden)

    Han Gyeol Song

    2013-07-01

    Full Text Available BackgroundRobots have allowed head and neck surgeons to extirpate oropharyngeal tumors safely without the need for lip-split incision or mandibulotomy. Using robots in oropharyngeal reconstruction is new but essential for oropharyngeal defects that result from robotic tumor excision. We report our experience with robotic free-flap reconstruction of head and neck defects to exemplify the necessity for robotic reconstruction.MethodsWe investigated head and neck cancer patients who underwent ablation surgery and free-flap reconstruction by robot. Between July 1, 2011 and March 31, 2012, 5 cases were performed and patient demographics, location of tumor, pathologic stage, reconstruction methods, flap size, recipient vessel, necessary pedicle length, and operation time were investigated.ResultsAmong five free-flap reconstructions, four were radial forearm free flaps and one was an anterolateral thigh free-flap. Four flaps used the superior thyroid artery and one flap used a facial artery as the recipient vessel. The average pedicle length was 8.8 cm. Flap insetting and microanastomosis were achieved using a specially manufactured robotic instrument. The total operation time was 1,041.0 minutes (range, 814 to 1,132 minutes, and complications including flap necrosis, hematoma, and wound dehiscence did not occur.ConclusionsThis study demonstrates the clinically applicable use of robots in oropharyngeal reconstruction, especially using a free flap. A robot can assist the operator in insetting the flap at a deep portion of the oropharynx without the need to perform a traditional mandibulotomy. Robot-assisted reconstruction may substitute for existing surgical methods and is accepted as the most up-to-date method.

  18. Modelling of in-vessel retention after relocation of corium into the lower plenum - Evaluation of the temperature field and of the viscoplastic deformation of the vessel wall. Reactor safety research, project No.:150 1254 - Final report; Beitrag zur Modellierung der Schmelzerueckhaltung im RDB nach Verlagerung von Corium in das untere Plenum - Berechnung des Temperaturfeldes und der viskoplastischen Verformung der Behaelterwand. Reaktorsicherheitsforschung, Vorhaben-Nr.: 150 1254 - Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Altstadt, E.; Willschuetz, H.G. [Forschungszentrum Rossendorf e.V. (FZR), Dresden (Germany)

    2005-01-01

    Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute Of Safety Research of the FZR a finite element model has been developed simulating the thermal processes and the viscoplastic behaviour of the vessel wall. An advanced model for creep and material damage has been established and has been validated using experimental data. The thermal hydraulic and the mechanical calculations are sequentially and recursively coupled. The model is capable of evaluating fracture time and fracture position of a vessel with an internally heated melt pool. The model was applied to pre- and post test calculations for the FOREVER test series representing the RPV of a PWR in the scale of 1:10. These experiments were performed at the Royal Institute of Technology in Stockholm. The results of the calculations can be summarised as follows: The creeping process is caused by the simultaneous presence of high temperature (>600 C) and pressure (>1 MPa). The hot focus region is the most endangered zone exhibiting the highest creep strain rates. The exact level of temperature and pressure has an influence on the vessel failure time but not on the failure position. The failure time can be predicted with an uncertainty of 20 to 25%. This uncertainty is caused by the large scatter and the high temperature sensitivity of the viscoplastic properties of the RPV steel. Contrary to the hot focus region, the lower centre of the vessel head exhibits a higher strength because of the lower temperatures in this zone. The lower part moves down without significant deformation. Therefore it can be assumed, that the vessel failure can be retarded or prevented by supporting this range. The

  19. 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.

  20. Integrity of PWR pressure vessels during overcooling accidents

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

    1982-01-01

    The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. For the purpose of evaluating this problem a state-of-the-art fracture mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure today if subjected to a Rancho Seco (1978) or TMI-2 (1979) type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

  1. Integrity of PWR pressure vessels during overcooling accidents

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

    1982-01-01

    The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. A state-of-the-art fracture-mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure in a few years if subjected to a Rancho Seco-type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

  2. Assessing women's lacrosse head impacts using finite element modelling.

    Science.gov (United States)

    Clark, J Michio; Hoshizaki, T Blaine; Gilchrist, Michael D

    2018-04-01

    Recently studies have assessed the ability of helmets to reduce peak linear and rotational acceleration for women's lacrosse head impacts. However, such measures have had low correlation with injury. Maximum principal strain interprets loading curves which provide better injury prediction than peak linear and rotational acceleration, especially in compliant situations which create low magnitude accelerations but long impact durations. The purpose of this study was to assess head and helmet impacts in women's lacrosse using finite element modelling. Linear and rotational acceleration loading curves from women's lacrosse impacts to a helmeted and an unhelmeted Hybrid III headform were input into the University College Dublin Brain Trauma Model. The finite element model was used to calculate maximum principal strain in the cerebrum. The results demonstrated for unhelmeted impacts, falls and ball impacts produce higher maximum principal strain values than stick and shoulder collisions. The strain values for falls and ball impacts were found to be within the range of concussion and traumatic brain injury. The results also showed that men's lacrosse helmets reduced maximum principal strain for follow-through slashing, falls and ball impacts. These findings are novel and demonstrate that for high risk events, maximum principal strain can be reduced by implementing the use of helmets if the rules of the sport do not effectively manage such situations. Copyright © 2018 Elsevier Ltd. All rights reserved.

  3. Maury Journals - US Vessels

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — U.S. vessels observations, after the 1853 Brussels Conference that set International Maritime Standards, modeled after Maury Marine Standard Observations.

  4. Use of miniature and standard specimens to evaluate effects of irradiation temperature on pressure vessel steels

    International Nuclear Information System (INIS)

    Haggag, F.M.; Nanstad, R.K.; Byrne, S.T.

    1991-01-01

    The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low [121 to 210 degrees C (250--410 degrees F)] compared to those for commercial light-water reactors (LWRs) [∼288 degrees C (550 degrees F)]. The need for design data on the reference temperature shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plate. A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, the vessel flange. This paper presents results from the first four irradiation capsules of this program. The four capsules were irradiated in the University of Buffalo Reactor to an effective fast fluence of 1 x10 18 neutron/cm 2 [0.68 x 10 18 neutron/cm 2 (>1 MeV)] at temperatures of 288, 204, 163, and 121 degrees C (550, 400, 325, and 250 degrees F), respectively. The yield and ultimate strengths of both steel plate materials of the MHTGR Program increased with decreasing irradiation temperature. Similarly, the 41-J Charpy V-notch (CVN) transition temperature shift increased with decreasing irradiation temperature (in agreement with the increase in yield strength). The miniature tensile and automated ball indentation (ABI) test results (yield strength and flow properties) were in good agreement with those from standard tensile specimens. The miniature tensile and ABI test results were also used in a model that utilizes the changes in yield strength to estimate the CVN ductile-to-brittle transition temperature shift due to irradiation. The model predictions were compared with CVN test results obtained here and in earlier work. 5 refs., 11 figs., 6 tabs

  5. Failure prediction of low-carbon steel pressure vessel and cylindrical models

    International Nuclear Information System (INIS)

    Zhang, K.D.; Wang, W.

    1987-01-01

    The failure loads predicted by failure assessment methods (namely the net-section stress criterion; the EPRI engineering approach for elastic-plastic analysis; the CEGB failure assessment route; the modified R6 curve by Milne for strain hardening; and the failure assessment curve based on J estimation by Ainsworth) have been compared with burst test results on externally, axially sharp notched pressure vessel and open-ended cylinder models made from typical low-carbon steel St45 seamless tube which has a transverse true stress-strain curve of straight-line and parabola type and a high value of ultimate strength to yield. It was concluded from the comparison that whilst the net-section stress criterion and the CEGB route did not give conservative predictions, Milne's modified curve did give a conservative and good prediction; Ainsworth's curve gave a fairly conservative prediction; and EPRI solutions also could conditionally give a good prediction but the conditions are still somewhat uncertain. It is suggested that Milne's modified R6 curve is used in failure assessment of low-carbon steel pressure vessels. (author)

  6. Basic Boiling Experiments with An Inclined Narrow Gap Associated With In-Vessel Retention

    International Nuclear Information System (INIS)

    Terazu, Kuninobu; Watanabe, Fukashi; Iwaki, Chikako; Yokobori, Seiichi; Akinaga, Makoto; Hamazaki, Ryoichi; SATO, Ken-ichi

    2002-01-01

    In the case of a severe accident with relocation of the molten corium into the lower plenum of reactor pressure vessel (RPV), the successful in-vessel corium retention (IVR) can prevent the progress to ex-vessel events with uncertainties and avoid the containment failure. One of the key phenomena governing the possibility of IVR would be the gap formation and cooling between a corium crust and the RPV wall, and for the achievement of IVR, it would be necessary to supply cooling water to RPV as early as possible. The BWR features relative to IVR behavior are a deep and massive water pool in the lower plenum, and many of control rod drive guide tubes (CRDGT) installed in the lower head of RPV, in which water is injected continuously except in the case of station blackout scenario. The present paper describes the basic boiling experiment conducted in order to investigate the boiling characteristics in an inclined narrow gap simulating a part of the lower head curvature. The boiling experiments were composed of visualization tests and heat transfer tests. In the visualization tests, two types of inclined gap were constructed using the parallel plate and the V-shaped parallel plate with heating from the top plate, and the boiling flow pattern was observed with various gap width and heat flux. These observation results showed that water was easily supplied from the gap bottom of parallel plate even in a very narrow gap with smaller width than 1 mm, and water could flow continuously in the narrow gap by the geometric and thermal imbalance from the experiment results using the V-shaped parallel plate. In the heat transfer tests, the critical heat flux (CHF) data in an inclined narrow channel formed by the parallel plates were measured in terms of the parameters of gap width, heated length and inclined angle of a channel, and the effect of inclination was incorporated into the existing CHF correlation for a narrow gap. The CHF correlation modified for an inclined narrow gap

  7. Green vessel scheduling in liner shipping: Modeling carbon dioxide emission costs in sea and at ports of call

    Directory of Open Access Journals (Sweden)

    Maxim A. Dulebenets

    2018-03-01

    Full Text Available Considering a substantial increase in volumes of the international seaborne trade and drastic climate changes due to carbon dioxide emissions, liner shipping companies have to improve planning of their vessel schedules and improve energy efficiency. This paper presents a novel mixed integer non-linear mathematical model for the green vessel scheduling problem, which directly accounts for the carbon dioxide emission costs in sea and at ports of call. The original non-linear model is linearized and then solved using CPLEX. A set of numerical experiments are conducted for a real-life liner shipping route to reveal managerial insights that can be of importance to liner shipping companies. Results indicate that the proposed mathematical model can serve as an efficient planning tool for liner shipping companies and may assist with evaluation of various carbon dioxide taxation schemes. Increasing carbon dioxide tax may substantially change the design of vessel schedules, incur additional route service costs, and improve the environmental sustainability. However, the effects from increasing carbon dioxide tax on the marine container terminal operations are found to be very limited.

  8. Vessel size measurements in angiograms: A comparison of techniques

    International Nuclear Information System (INIS)

    Hoffmann, Kenneth R.; Nazareth, Daryl P.; Miskolczi, Laszlo; Gopal, Anant; Wang Zhou; Rudin, Stephen; Bednarek, Daniel R.

    2002-01-01

    As interventional procedures become more complicated, the need for accurate quantitative vascular information increases. In response to this need, many commercial vendors provide techniques for measurement of vessel sizes, usually based on derivative techniques. In this study, we investigate the accuracy of several techniques used in the measurement of vessel size. Simulated images of vessels having circular cross sections were generated and convolved with various focal spot distributions taking into account the magnification. These vessel images were then convolved with Gaussian image detector line spread functions (LSFs). Additionally, images of a phantom containing vessels with a range of diameters were acquired for the 4.5'', 6'', 9'', and 12'' modes of an image intensifier-TV (II-TV) system. Vessel sizes in the images were determined using a first-derivative technique, a second-derivative technique, a linear combination of these two measured sizes, a thresholding technique, a densitometric technique, and a model-based technique. For the same focal spot size, the shape of the focal spot distribution does not affect measured vessel sizes except at large magnifications. For vessels with diameters larger than the full-width-at-half-maximum (FWHM) of the LSF, accurate vessel sizes (errors ∼0.1 mm) could be obtained by using an average of sizes determined by the first and second derivatives. For vessels with diameters smaller than the FWHM of the LSF, the densitometric and model-based techniques can provide accurate vessel sizes when these techniques are properly calibrated

  9. A comparison of elastic-plastic and variable modulus-cracking constitutive models for prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Anderson, C.A.; Smith, P.D.

    1979-01-01

    Numerical prediction of the behavior of prestressed concrete reactor vessels (PCRVs) under static, dynamic and long term loadings is complicated by the currently ill-defined behavior of concrete under stress and the three-dimensional nature of PCRVs. Which constitutive model most closely approximates the behavior of concrete in PCRVs under load has not yet been decided. Many equations for accurately modeling the three-dimensional behavior of PCRVs tax the capability of a most up-to-date computing system. The main purpose of this paper is to compare the characteristics of two constitutive models which have been proposed for concrete, variable modulus cracking model and elastic-plastic model. Moreover, the behavior of typical concrete structures was compared, the materials of which obey these constitutive laws. The response to internal pressure of PCRV structure, the constitutive models for concrete, the test problems using a thick-walled concrete ring and a rectangular concrete plate, and the analysis of an axisymmetric concrete pressure vessel PV-26 using the variable modulus cracking model of the ADINA code are explained. The variable modulus cracking model can predict the behavior of reinforced concrete structures well into the range of nonlinear behavior. (Kako, I.)

  10. A MODEL OF MIRA'S COMETARY HEAD/TAIL ENTERING THE LOCAL BUBBLE

    International Nuclear Information System (INIS)

    Esquivel, A.; Raga, A. C.; RodrIguez-Gonzalez, A.; Lopez-Camara, D.; Velazquez, P. F.; Canto, J.; De Colle, F.

    2010-01-01

    We model the cometary structure around Mira as the interaction of an asymptotic giant branch stellar wind from Mira A with a streaming environment. Our simulations introduce the following new element: we assume that after 200 kyr of evolution in a dense environment, Mira entered the Local Bubble (low-density coronal gas). As Mira enters the bubble, the head of the comet expands quite rapidly, while the tail remains well collimated for a >100 kyr timescale. The result is a broad-head/narrow-tail structure that resembles the observed morphology of Mira's comet. The simulations were carried out with our new adaptive grid code WALICXE, which is described in detail.

  11. Heatup of the TMI-2 lower head during core relocation

    International Nuclear Information System (INIS)

    Wang, S.K.; Sienicki, J.J.; Spencer, B.W.

    1989-01-01

    An analysis has been carried out to assess the potential of a melting attack upon the reactor vessel lower head and incore instrument nozzle penetration weldments during the TMI core relocation event at 224 minutes. Calculations were performed to determine the potential for molten corium to undergo breakup into droplets which freeze and form a debris bed versus impinging upon the lower head as one or more coherent streams. The effects of thermal-hydraulic interactions between corium streams and water inside the lower plenum, the effects of the core support assembly structure upon the corium, and the consequences of corium relocation by way of the core former region were examined. 19 refs., 24 figs

  12. OceanRoute: Vessel Mobility Data Processing and Analyzing Model Based on MapReduce

    Science.gov (United States)

    Liu, Chao; Liu, Yingjian; Guo, Zhongwen; Jing, Wei

    2018-06-01

    The network coverage is a big problem in ocean communication, and there is no low-cost solution in the short term. Based on the knowledge of Mobile Delay Tolerant Network (MDTN), the mobility of vessels can create the chances of end-to-end communication. The mobility pattern of vessel is one of the key metrics on ocean MDTN network. Because of the high cost, few experiments have focused on research of vessel mobility pattern for the moment. In this paper, we study the traces of more than 4000 fishing and freight vessels. Firstly, to solve the data noise and sparsity problem, we design two algorithms to filter the noise and complement the missing data based on the vessel's turning feature. Secondly, after studying the traces of vessels, we observe that the vessel's traces are confined by invisible boundary. Thirdly, through defining the distance between traces, we design MR-Similarity algorithm to find the mobility pattern of vessels. Finally, we realize our algorithm on cluster and evaluate the performance and accuracy. Our results can provide the guidelines on design of data routing protocols on ocean MDTN.

  13. Improved transcranial magnetic stimulation coil design with realistic head modeling

    Science.gov (United States)

    Crowther, Lawrence; Hadimani, Ravi; Jiles, David

    2013-03-01

    We are investigating Transcranial magnetic stimulation (TMS) as a noninvasive technique based on electromagnetic induction which causes stimulation of the neurons in the brain. TMS can be used as a pain-free alternative to conventional electroconvulsive therapy (ECT) which is still widely implemented for treatment of major depression. Development of improved TMS coils capable of stimulating subcortical regions could also allow TMS to replace invasive deep brain stimulation (DBS) which requires surgical implantation of electrodes in the brain. Our new designs allow new applications of the technique to be established for a variety of diagnostic and therapeutic applications of psychiatric disorders and neurological diseases. Calculation of the fields generated inside the head is vital for the use of this method for treatment. In prior work we have implemented a realistic head model, incorporating inhomogeneous tissue structures and electrical conductivities, allowing the site of neuronal activation to be accurately calculated. We will show how we utilize this model in the development of novel TMS coil designs to improve the depth of penetration and localization of stimulation produced by stimulator coils.

  14. Predicting Vessel Trajectories from Ais Data Using R

    Science.gov (United States)

    2017-06-01

    Source: Hampton (2009). A vessel operator with AIS is able to get useful information about the other vessels in the area by selecting a vessel icon ...random forest model on our computer. All calculations are done on a MacBook-Pro with 2.7GHz quad-core Intel Core i7, and 16GB of memory . H2O allows us

  15. Head-facial hemangiomas studied with scanning electron microscopy.

    Science.gov (United States)

    Cavallotti, Carlo; Cavallotti, Chiara; Giovannetti, Filippo; Iannetti, Giorgio

    2009-11-01

    Hemangiomas of the head or face are a frequent vascular pathology, consisting in an embryonic dysplasia that involves the cranial-facial vascular network. Hemangiomas show clinical, morphological, developmental, and structural changes during their course. Morphological, structural, ultrastructural, and clinical characteristics of head-facial hemangiomas were studied in 28 patients admitted in our hospital. Nineteen of these patients underwent surgery for the removal of the hemangiomas, whereas 9 patients were not operated on. All the removed tissues were transferred in our laboratories for the morphological staining. Light microscopy, transmission electron microscopy, and scanning electron microscopy techniques were used for the observation of all microanatomical details. All patients were studied for a clinical diagnosis, and many were subjected to surgical therapy. The morphological results revealed numerous microanatomical characteristics of the hemangiomatous vessels. The observation by light microscopy shows the afferent and the efferent vessels for every microhemangioma. All the layers of the arterial wall are uneven. The lumen of the arteriole is entirely used by a blood clot. The observation by transmission electron microscopy shows that it was impossible to see the limits of the different layers (endothelium, medial layer, and adventitia) in the whole wall of the vessels. Moreover, both the muscular and elastic components are disarranged and replaced with connective tissue. The observation by scanning electron microscopy shows that the corrosion cast of the hemangioma offers 3 periods of filling: initially with partial filling of the arteriolar and of the whole cast, intermediate with the entire filling of the whole cast (including arteriole and venule), and a last period with a partial emptying of the arteriolar and whole cast while the venule remains totally injected with resin. Our morphological results can be useful to clinicians for a precise

  16. Development of heat transfer enhancement techniques for external cooling of an advanced reactor vessel

    Science.gov (United States)

    Yang, Jun

    Nucleate boiling is a well-recognized means for passively removing high heat loads (up to ˜106 W/m2) generated by a molten reactor core under severe accident conditions while maintaining relatively low reactor vessel temperature (Critical Heat Flux (CHF), becomes the key to the success of external passive cooling of reactor vessel undergoing core disrupture accidents. In the present study, two boiling heat transfer enhancement methods have been proposed, experimentally investigated and theoretically modelled. The first method involves the use of a suitable surface coating to enhance downward-facing boiling rate and CHF limit so as to substantially increase the possibility of reactor vessel surviving high thermal load attack. The second method involves the use of an enhanced vessel/insulation design to facilitate the process of steam venting through the annular channel formed between the reactor vessel and the insulation structure, which in turn would further enhance both the boiling rate and CHF limit. Among the various available surface coating techniques, metallic micro-porous layer surface coating has been identified as an appropriate coating material for use in External Reactor Vessel Cooling (ERVC) based on the overall consideration of enhanced performance, durability, the ease of manufacturing and application. Since no previous research work had explored the feasibility of applying such a metallic micro-porous layer surface coating on a large, downward facing and curved surface such as the bottom head of a reactor vessel, a series of characterization tests and experiments were performed in the present study to determine a suitable coating material composition and application method. Using the optimized metallic micro-porous surface coatings, quenching and steady-state boiling experiments were conducted in the Sub-scale Boundary Layer Boiling (SBLB) test facility at Penn State to investigate the nucleate boiling and CHF enhancement effects of the surface

  17. A two dimensional approach for temperature distribution in reactor lower head during severe accident

    International Nuclear Information System (INIS)

    Cao, Zhen; Liu, Xiaojing; Cheng, Xu

    2015-01-01

    Highlights: • Two dimensional module is developed to analyze integrity of lower head. • Verification step has been done to evaluate feasibility of new module. • The new module is applied to simulate large-scale advanced PWR. • Importance of 2-D approach is clearly quantified. • Major parameters affecting vessel temperature distribution are identified. - Abstract: In order to evaluate the safety margin during a postulated severe accident, a module named ASAP-2D (Accident Simulation on Pressure vessel-2 Dimensional), which can be implemented into the severe accident simulation codes (such as ATHLET-CD), is developed in Shanghai Jiao Tong University. Based on two-dimensional spherical coordinates, heat conduction equation for transient state is solved implicitly. Together with solid vessel thickness, heat flux distribution and heat transfer coefficient at outer vessel surface are obtained. Heat transfer regime when critical heat flux has been exceeded (POST-CHF regime) could be simulated in the code, and the transition behavior of boiling crisis (from spatial and temporal points of view) can be predicted. The module is verified against a one-dimensional analytical solution with uniform heat flux distribution, and afterwards this module is applied to the benchmark illustrated in NUREG/CR-6849. Benchmark calculation indicates that maximum heat flux at outer surface of RPV could be around 20% lower than that of at inner surface due to two-dimensional heat conduction. Then a preliminary analysis is performed on the integrity of the reactor vessel for which the geometric parameters and boundary conditions are derived from a large scale advanced pressurized water reactor. Results indicate that heat flux remains lower than critical heat flux. Sensitivity analysis indicates that outer heat flux distribution is more sensitive to input heat flux distribution and the transition boiling correlation than mass flow rate in external reactor vessel cooling (ERVC) channel

  18. Modeling with finite element of the upper head spring; Modelizacion con elementos finitos del resorte del cabezal superior

    Energy Technology Data Exchange (ETDEWEB)

    Munoz Cardador, J.; Cerrain Arranz, A.

    2013-07-01

    The objective of this work is the development of a model of finite element of the upper head spring so that it can be used as a tool in the design of the same. For this purpose, simulates the behavior to compression spring of the integrated head 17 x 17 using a numerical model and are validated with experimental results obtained in tests conducted by ENUSA. The validated model is a new tool to the spring design of the upper head whose use can extend both for the evaluation of current designs as for the evaluation of new modifications.

  19. A mathematical model for pressure-based organs behaving as biological pressure vessels.

    Science.gov (United States)

    Casha, Aaron R; Camilleri, Liberato; Gauci, Marilyn; Gatt, Ruben; Sladden, David; Chetcuti, Stanley; Grima, Joseph N

    2018-04-26

    We introduce a mathematical model that describes the allometry of physical characteristics of hollow organs behaving as pressure vessels based on the physics of ideal pressure vessels. The model was validated by studying parameters such as body and organ mass, systolic and diastolic pressures, internal and external dimensions, pressurization energy and organ energy output measurements of pressure-based organs in a wide range of mammals and birds. Seven rules were derived that govern amongst others, lack of size efficiency on scaling to larger organ sizes, matching organ size in the same species, equal relative efficiency in pressurization energy across species and direct size matching between organ mass and mass of contents. The lung, heart and bladder follow these predicted theoretical relationships with a similar relative efficiency across various mammalian and avian species; an exception is cardiac output in mammals with a mass exceeding 10kg. This may limit massive body size in mammals, breaking Cope's rule that populations evolve to increase in body size over time. Such a limit was not found in large flightless birds exceeding 100kg, leading to speculation about unlimited dinosaur size should dinosaurs carry avian-like cardiac characteristics. Copyright © 2018. Published by Elsevier Ltd.

  20. Modeling the Role of the Glymphatic Pathway and Cerebral Blood Vessel Properties in Alzheimer's Disease Pathogenesis.

    Directory of Open Access Journals (Sweden)

    Christina Rose Kyrtsos

    Full Text Available Alzheimer's disease (AD is the most common cause of dementia in the elderly, affecting over 10% population over the age of 65 years. Clinically, AD is described by the symptom set of short term memory loss and cognitive decline, changes in mentation and behavior, and eventually long-term memory deficit as the disease progresses. On imaging studies, significant atrophy with subsequent increase in ventricular volume have been observed. Pathology on post-mortem brain specimens demonstrates the classic findings of increased beta amyloid (Aβ deposition and the presence of neurofibrillary tangles (NFTs within affected neurons. Neuroinflammation, dysregulation of blood-brain barrier transport and clearance, deposition of Aβ in cerebral blood vessels, vascular risk factors such as atherosclerosis and diabetes, and the presence of the apolipoprotein E4 allele have all been identified as playing possible roles in AD pathogenesis. Recent research has demonstrated the importance of the glymphatic system in the clearance of Aβ from the brain via the perivascular space surrounding cerebral blood vessels. Given the variety of hypotheses that have been proposed for AD pathogenesis, an interconnected, multilayer model offers a unique opportunity to combine these ideas into a single unifying model. Results of this model demonstrate the importance of vessel stiffness and heart rate in maintaining adequate clearance of Aβ from the brain.

  1. Investigation of the Potential for In-Vessel Melt Retention in the Lower Head of a BWR by Cooling through the Control Rod Guide Tubes. APRl 4, Stage 2 Report

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Jasiulevicius, A.; Konovalikhin, M.

    2004-01-01

    recommended that further investigations, both experimental and model development, be conducted to (a) check reproducibility of data (b) employ different flow rates (c) employ different simulant materials and (d) develop a comprehensive model, in order to certify that the coolability that can be achieved with establishing a water flow in the CRGTs will be able to retain the melt in the lower head of a BWR. We believe it will be an extremely important accident management strategy for a Swedish BWR since it will obviate the consideration of the prime licensing issue of ex-vessel steam explosion induced containment failure associated with the present scheme of establishing a water pool in the lower drywell of all the Swedish BWRs

  2. Shortened outage duration and increased safety with head assembly upgrade packages

    International Nuclear Information System (INIS)

    Leanne, M.; Lisien, P.E.; Plute, K.; Duran, J.

    2007-01-01

    To significantly reduce outage critical path duration and personnel radiation exposure, and to increase personnel safety, Westinghouse Electric Co., LLC has designed and installed upgrades to the existing head assemblies of operating pressurized water reactors. These upgrades are known as Head Assembly Upgrade Packages (HAUPs) or Simplified Head Assemblies (SHAs). Custom configurations are created from a set of standard elements to optimize the design for each unique containment, head assembly configuration, and licensing basis. Two primary options are available for implementation: a full HAUP or targeted component and system upgrades. Plants may achieve much of the outage savings, dose reduction, and safety improvements even with a more limited hardware scope. A range of improvements can be offered from integral missile shields, to redesigned duct work, radiation shields, and cable layout and connection optimization. The hardware changes are customized to target the scope that adds the most value for a given plant. While combining upgrades with a reactor vessel head (RVH) replacement adds some flexibility, it is not necessary. Some plants have chosen to implement targeted upgrades prior to a replacement RVH outage and then complete the remainder of the full HAUP during the replacement RVH outage. Three-dimensional computer aided design tools are used in the conceptual and detailed design phases to identify and avoid interferences between existing and replacement plant components. State-of-the-art computational fluid dynamics models for control drive mechanism (CDM) cooling systems are used to demonstrate the ability to maintain or improve the original design performance while greatly simplifying the disassembly/re-assembly activities. Likewise, state-of-the-art finite element analysis methods allow optimization of structural components while meeting code limits for design basis accident conditions. (authors)

  3. Estimation on the Flow Phenomena and the Pressure Loss for the Inlet Part of a Research Reactor Vessel

    International Nuclear Information System (INIS)

    Seo, Kyoung Woo; Oh, Jae Min; Seo, Jae Kwang; Yoon, Ju Hyeon; Lee, Doo Jeong

    2009-01-01

    For a research reactor, a conceptual primary cooling system (PCS) was designed for an adequate cooling to the reactor core. The developed primary cooling circuit consisted of decay tanks, pumps, heat exchangers, vacuum breakers, some isolation and check valves, connection piping, and instruments. The main function of the primary cooling pumps (PCPs) of the PCS was to circulate the reactor coolant through the fuel core and the heat exchangers during a normal operation. The head according to the design flow rate which was determined by the thermal hydraulic design analysis for the core should be estimated to design the PCPs in the fluid system. The pressure loss in the PCS can be calculated by the dimensional analysis of the pipe flow and the head loss coefficient of the components. However, it is insufficient to estimate the pressure loss for 3-dimensional flow phenomena such as the flow path in the reactor with the theoretical dimensional analysis based on experimental data. The purpose of this research is to evaluate the pressure loss of the part of a research reactor vessel. For evaluating the pressure loss, the commercially available CFD computer model, FLUENT, was employed. First, for validating the application of FLUENT to the pressure loss, a simple case was calculated and compared with the Idelchik empirical correlation. Secondly, several cases for the inlet part of a research reactor vessel were estimated by a FLUENT 3- dimensional calculation

  4. Right thalamic infarction after closed head injury

    International Nuclear Information System (INIS)

    Nagaya, Takashi; Doi, Terushige; Katsumata, Tsuguo; Kuwayama, Naoto

    1986-01-01

    We reported a case of right thalamic infarction after a closed head injury. A 12-year-old boy was hit by an autotruck. He was semi-comatose, with left temporal scalp swelling and excoriation in the left lower limb. Three days after the accident, he exhibited left hemiparesis. CT scans on the day of the accident showed no abnormality, but on the following day, right thalamic infarction appeared. Right carotid angiography showed only an irregular vascular shadow in the cisternal segment of the right internal carotid artery. Vascular obstruction after closed head injury is rare, especially in the intracranial vessels, and several pathogeneses may be postulated. The right thalamic infarction in this case was supposed to be due to the damage of the perforators from the right posterior communicating artery and the right posterior cerebral artery, which were struck as a contre-coup by the force from the left side. (author)

  5. Analysis code for pressure in reactor containment vessel of ATR. CONPOL

    International Nuclear Information System (INIS)

    1997-08-01

    For the evaluation of the pressure and temperature in containment vessels in the events which are classified in the abnormal change of pressure, atmosphere and others in reactor containment vessels in accident among the safety evaluation events of the ATR, the analysis code for the pressure in reactor containment vessels CONPOL is used. In this report, the functions of the analysis code and the analysis model are shown. By using this analysis code, the rise of the pressure and temperature in a containment vessel is evaluated when loss of coolant accident occurs, and high temperature, high pressure coolant flows into it. This code possesses the functions of computing blow-down quantity and heat dissipation from reactor cooling facility, steam condensing heat transfer to containment vessel walls, and the cooling effect by containment vessel spray system. As for the analysis techniques, the models of reactor cooling system, containment vessel and steam discharge pool, and the computation models for the pressure and temperature in containment vessels, wall surface temperature, condensing heat transfer, spray condensation and blow-down are explained. The experimental analysis of the evaluation of the pressure and temperature in containment vessels at the time of loss of coolant accident is reported. (K.I.)

  6. Diffusive oxygen shunting between vessels in the preglomerular renal vasculature: anatomic observations and computational modeling.

    Science.gov (United States)

    Gardiner, Bruce S; Thompson, Sarah L; Ngo, Jennifer P; Smith, David W; Abdelkader, Amany; Broughton, Brad R S; Bertram, John F; Evans, Roger G

    2012-09-01

    To understand how geometric factors affect arterial-to-venous (AV) oxygen shunting, a mathematical model of diffusive oxygen transport in the renal cortex was developed. Preglomerular vascular geometry was investigated using light microscopy (providing vein shape, AV separation, and capillary density near arteries) and published micro-computed tomography (CT) data (providing vessel size and AV separation; Nordsletten DA, Blackett S, Bentley MD, Ritman EL, Smith NP. IUPS Physiome Project. http://www.physiome.org.nz/publications/nordsletten_blackett_ritman_bentley_smith_2005/folder_contents). A "U-shaped" relationship was observed between the arterial radius and the distance between the arterial and venous lumens. Veins were found to partially wrap around the artery more consistently for larger rather than smaller arteries. Intrarenal arteries were surrounded by an area of fibrous tissue, lacking capillaries, the thickness of which increased from ∼5 μm for the smallest arteries (200-μm diameter). Capillary density was greater near smaller arteries than larger arteries. No capillaries were observed between wrapped AV vessel pairs. The computational model comprised a single AV pair in cross section. Geometric parameters critical in renal oxygen transport were altered according to variations observed by CT and light microscopy. Lumen separation and wrapping of the vein around the artery were found to be the critical geometric factors determining the amount of oxygen shunted between AV pairs. AV oxygen shunting increases both as lumen separation decreases and as the degree of wrapping increases. The model also predicts that capillaries not only deliver oxygen, but can also remove oxygen from the cortical parenchyma close to an AV pair. Thus the presence of oxygen sinks (capillaries or tubules) near arteries would reduce the effectiveness of AV oxygen shunting. Collectively, these data suggest that AV oxygen shunting would be favored in larger vessels common to the

  7. FE-simulation of the viscoplastic behaviour of different RPV steels in the frame of in-vessel melt retentions scenarios

    International Nuclear Information System (INIS)

    Altstadt, E.; Willschuetz, H.G.; Mueller, G.

    2004-01-01

    Assuming the hypothetical scenario of a severe accident with subsequent core meltdown and formation of a melt pool in the reactor pressure vessel (RPV) lower plenum of a Light Water Reactor (LWR) leads to the question about the behavior of the RPV. One accident management strategy could be to stabilize the in-vessel debris configuration in the RPV as one major barrier against uncontrolled release of heat and radio nuclides. To get an improved understanding and knowledge of the melt pool convection and the vessel creep and possible failure processes and modes occurring during the late phase of a core melt down accident the FOREVER-experiments (Failure Of REactor VEssel Retention) have been performed at the Division of Nuclear Power Safety of the Royal Institute of Technology Stockholm. These experiments are simulating the behavior of the lower head of the RPV under the thermal loads of a convecting melt pool with decay heating, and under the pressure loads that the vessel experiences in a depressurization scenario. The geometrical scale of the experiments is 1:10 compared to a common LWR. This paper deals with the experimental, numerical, and metallographical results of the creep failure experiment EC-FOREVER-4, where the American pressure vessel steel SA533B was applied for the lower head. For comparison the results of the experiment EC-FOREVER-3B, build of the French 16MND5 steel, are discussed, too. Emphasis is put on the differences in the viscoplastic behaviour of different heats of the RPV steel. For this purpose, the creep tests in the frame of the LHF/OLHF experiments are reviewed, too. As a hypothesis it is stated that the sulphur content could be responsible for differences in the creep behaviour. (orig.)

  8. Development of FB-MultiPier dynamic vessel-collision analysis models, phase 2 : [summary].

    Science.gov (United States)

    2014-07-01

    When collisions between large vessels and bridge : supports occur, they can result in significant : damage to bridge and vessel. These collisions : are extremely hazardous, often taking lives on : the vessel and the bridge. Direct costs of repair : a...

  9. Popliteal vascular entrapment syndrome caused by a rare anomalous slip of the lateral head of the gastrocnemius muscle

    International Nuclear Information System (INIS)

    Liu, Patrick T.; Moyer, Adrian C.; Huettl, Eric A.; Fowl, Richard J.; Stone, William M.

    2005-01-01

    Popliteal vascular entrapment syndrome can result in calf claudication, aneurysm formation, distal arterial emboli, or popliteal vessel thrombosis. The most commonly reported causes of this syndrome have been anomalies of the medial head of the gastrocnemius muscle as it relates to the course of the popliteal artery. We report two cases of rare anomalous slips of the lateral head of the gastrocnemius muscle causing popliteal vascular entrapment syndrome. (orig.)

  10. Spectra and neutron dose of an 18 MV Linac using two geometric models of the head

    International Nuclear Information System (INIS)

    Barrera, M. T.; Pino, F.; Barros, H.; Sajo-Bohus, L.; Davila, J.; Salcedo, E.; Vega C, H. R.; Benites R, J. L.

    2015-10-01

    Full text: Using the Monte Carlo method, by MCNP5 code, simulations were performed with different source terms and 2 geometric models of the head to obtain spectra in energy, flow and doses of photo-neutrons at different positions on the stretcher and in the radiotherapy room. The simplest model was a spherical shell of tungsten; the second was the complete model of a heterogeneous head of an accelerator Varian ix. In both models Tosi function was used as a source term. In addition, for the second model Sheikh-Bagheri distribution was used for photons and photo-neutrons were generated. Also in both models the radiotherapy room of Gurve group of the Teaching Medical Center La Trinidad was included, which is equipped with an accelerator Varian Clinic 2100. In this Center passive detectors PADC (Cr-39) were irradiated with neutron converters, with 18 MeV photons radiation. The measured neutron flow was compared with that obtained with Monte Carlo calculations. The Monte Carlo flows are similar to those measured at the isocenter. The simplest model underestimates the neutron flow compared with the calculated flows with the heterogeneous model of the head. (Author)

  11. Modeling the Isentropic Head Value of Centrifugal Gas Compressor using Genetic Programming

    Directory of Open Access Journals (Sweden)

    Safiyullah Ferozkhan

    2016-01-01

    Full Text Available Gas compressor performance is vital in oil and gas industry because of the equipment criticality which requires continuous operations. Plant operators often face difficulties in predicting appropriate time for maintenance and would usually rely on time based predictive maintenance intervals as recommended by original equipment manufacturer (OEM. The objective of this work is to develop the computational model to find the isentropic head value using genetic programming. The isentropic head value is calculated from the OEM performance chart. Inlet mass flow rate and speed of the compressor are taken as the input value. The obtained results from the GP computational models show good agreement with experimental and target data with the average prediction error of 1.318%. The genetic programming computational model will assist machinery engineers to quantify performance deterioration of gas compressor and the results from this study will be then utilized to estimate future maintenance requirements based on the historical data. In general, this genetic programming modelling provides a powerful solution for gas compressor operators to realize predictive maintenance approach in their operations.

  12. Analysis of lower head failure with simplified models and a finite element code

    Energy Technology Data Exchange (ETDEWEB)

    Koundy, V. [CEA-IPSN-DPEA-SEAC, Service d' Etudes des Accidents, Fontenay-aux-Roses (France); Nicolas, L. [CEA-DEN-DM2S-SEMT, Service d' Etudes Mecaniques et Thermiques, Gif-sur-Yvette (France); Combescure, A. [INSA-Lyon, Lab. Mecanique des Solides, Villeurbanne (France)

    2001-07-01

    The objective of the OLHF (OECD lower head failure) experiments is to characterize the timing, mode and size of lower head failure under high temperature loading and reactor coolant system pressure due to a postulated core melt scenario. Four tests have been performed at Sandia National Laboratories (USA), in the frame of an OECD project. The experimental results have been used to develop and validate predictive analysis models. Within the framework of this project, several finite element calculations were performed. In parallel, two simplified semi-analytical methods were developed in order to get a better understanding of the role of various parameters on the creep phenomenon, e.g. the behaviour of the lower head material and its geometrical characteristics on the timing, mode and location of failure. Three-dimensional modelling of crack opening and crack propagation has also been carried out using the finite element code Castem 2000. The aim of this paper is to present the two simplified semi-analytical approaches and to report the status of the 3D crack propagation calculations. (authors)

  13. Proceedings of the IAEA specialists` meeting on cracking in LWR RPV head penetrations

    Energy Technology Data Exchange (ETDEWEB)

    Pugh, C.E.; Raney, S.J. [comps.] [Oak Ridge National Lab., TN (United States)

    1996-07-01

    This report contains 17 papers that were presented in four sessions at the IAEA Specialists` meeting on Cracking in LWR RPV Head Penetrations held at ASTM Headquarters in Philadelphia on May 2-3, 1995. The papers are compiled here in the order that presentations were made in the sessions, and they relate to operational observations, inspection techniques, analytical modeling, and regulatory control. The goal of the meeting was to allow international experts to review experience in the field of ensuring adequate performance of reactor pressure vessel (RPV) heads and penetrations. The emphasis was to allow a better understanding of RPV material behavior, to provide guidance supporting reliability and adequate performance, and to assist in defining directions for further investigations. The international nature of the meeting is illustrated by the fact that papers were presented by researchers from 10 countries. There were technical experts present form other countries who participated in discussions of the results presented. This present document incorporates the final version of the papers as received from the authors. The final chapter includes conclusions and recommendations. Individual papers have been cataloged separately.

  14. Proceedings of the IAEA specialists' meeting on cracking in LWR RPV head penetrations

    International Nuclear Information System (INIS)

    Pugh, C.E.; Raney, S.J.

    1996-07-01

    This report contains 17 papers that were presented in four sessions at the IAEA Specialists' meeting on Cracking in LWR RPV Head Penetrations held at ASTM Headquarters in Philadelphia on May 2-3, 1995. The papers are compiled here in the order that presentations were made in the sessions, and they relate to operational observations, inspection techniques, analytical modeling, and regulatory control. The goal of the meeting was to allow international experts to review experience in the field of ensuring adequate performance of reactor pressure vessel (RPV) heads and penetrations. The emphasis was to allow a better understanding of RPV material behavior, to provide guidance supporting reliability and adequate performance, and to assist in defining directions for further investigations. The international nature of the meeting is illustrated by the fact that papers were presented by researchers from 10 countries. There were technical experts present form other countries who participated in discussions of the results presented. This present document incorporates the final version of the papers as received from the authors. The final chapter includes conclusions and recommendations. Individual papers have been cataloged separately

  15. Neutron Assay System for Con?nement 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

    Waste will be removed from confinement vessels remaining from 1970s-era experiments. Los Alamos has 9+ spherical confinement vessels remaining from experiments. Each vessel contains ∼ 500 lbs of radioactive debris such as actinide metals and oxides, metals, powdered silica, graphite, and wires and hardware. In order to dispose of the vessels, debris and contamination must be removed. Neutron assay system was designed to assay vessels before and after cleanout. System requirements are: (1) Modular and moveable; (2) Capable of detecting ∼100g 239 Pu equivalent in a 2-inch thick steel sphere with 6 foot diameter; and (3) Capable of safeguards-quality assays. Initial design parameters arethe use of 4-atm 3 He tubes with length of 6 feet, and 3 He tubes embedded in polyethelene for moderation. This paper describes the calibration of the Confinement Vessel Assay System (CVAS) and quantification of its uncertainties. Assay uncertainty depends on five factors: (1) Statistical uncertainty in the assay measurement; (2) Statistical uncertainty in the background measurement; (3) Statistical uncertainty in the isotopics determination - This should be much smaller than the other uncertainties; (4) Systematic uncertainty due to position bias; and (5) Systematic uncertainty due to fluctuations in cosmic ray spallation. This one can be virtually eliminated by performing the background measurement with an empty vessel - but that may not be possible. We used modeling and experiments to quantify the systematic uncertainties. The calibration assumes a uniform distribution of material, but reality will be different. MCNPX modeling was used to quantify the positional bias. The model was benchmarked to build confidence in its results. Material at top of vessel is 44% greater than amount assayed, according to singles. Material near 19-tube detector is 38% less than amount assayed, according to singles. Cosmic ray spallation contributes significantly to the background. Comparing rates

  16. Simplified hydrodynamic model of hydrogen-flame propagation in reactor vessels

    International Nuclear Information System (INIS)

    Baer, M.R.; Ratzel, A.C.

    1983-01-01

    The model is consistent with the theory of slow combustion in which the gasdynamic field equations are treated in the limit of small Mach numbers. To the lowest order, pressure is spatially uniform. The flame is treated as a density and entropy discontinuity which propagates at prescribed burning velocities, corresponding to laminar or turbulent flames. Radiation cooling of the burned combustion gases and possible preheating of the unburned gases during propagation of the flame is included using a molecular gas-band thermal radiation model. Application of this model has been developed for 1-D variable-area flame propagation. Multidimensional effects induced by hydrodynamics and buoyancy are corrected for. This model of flame propagation reduces to differential equations which describes the temporal variations of vessel pressure, burned volume and gas entropy. The thermodynamic state of the burned gas immediately following the flame is determined using an isobaric Hugoniot relation. At other locations the burned-gas thermodynamic states are determined using a Lagrangian particle tracking method. Results of a computer code using the method are presented. 11 figures

  17. Reactor pressure vessel stud management automation strategies

    International Nuclear Information System (INIS)

    Biach, W.L.; Hill, R.; Hung, K.

    1992-01-01

    The adoption of hydraulic tensioner technology as the standard for bolting and unbolting the reactor pressure vessel (RPV) head 35 yr ago represented an incredible commitment to new technology, but the existing technology was so primitive as to be clearly unacceptable. Today, a variety of approaches for improvement make the decision more difficult. Automation in existing installations must meet complex physical, logistic, and financial parameters while addressing the demands of reduced exposure, reduced critical path, and extended plant life. There are two generic approaches to providing automated RPV stud engagement and disengagement: the multiple stud tensioner and automated individual tools. A variation of the latter would include the handling system. Each has its benefits and liabilities

  18. Design of the Intersector Welding Robot for vacuum vessel assembly and maintenance

    International Nuclear Information System (INIS)

    Jones, L.; Dagenais, J.-F.; Daenner, W.; Maisonnier, D.

    2000-01-01

    Next Step Fusion Devices require on-site (field weld) joining of sectors of the thick-walled vacuum vessel for structural and vacuum integrity. EFDA (European Fusion Development Agreement) is supporting an R and D programme to investigate processes for assembly of the vacuum vessel and to carry out cutting, re-welding and inspection for remote sector replacement, forming part of the overall VV/blanket research effort. In order to direct the process end-effectors along the field joint zone, a track-mounted Intersector Welding Robot (IWR) on a mock-up of a region of the vacuum vessel has been designed and is described in this paper. A rail-mounted hexapod type robot offers six axes of motion over a limited work envelope with high payload to robot weight ratio. A solution to the production of reduced pressure local vacuum is the installation of short, lightweight segments bolted to each other and the vessel wall. The various process heads can be mounted using end-effectors of special design. To minimise the supply and interface problems for the IWR prototype, its motion control and electronic systems will be embedded locally. A laser scan with camera forms the on-line seam tracking capability to compensate for rail and seam deviations

  19. Confinement Vessel Assay System: Design and Implementation Report

    International Nuclear Information System (INIS)

    Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Gomez, Cipriano D.; Miko, David K.; Salazar, William R.; Stange, Sy; Vigil, Georgiana M.

    2012-01-01

    Los Alamos National Laboratory has a number of spherical confinement vessels remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1- to 2-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. We have developed a neutron assay system for the purposes of Materials Control and Accountability (MC and A) measurements of the vessel prior to and after cleanout. We present our approach to confronting the challenges in designing, building, and testing such a system. The system was designed to meet a set of functional and operational requirements. A Monte Carlo model was developed to aid in optimizing the detector design as well as to predict the systematic uncertainty associated with confinement vessel measurements. Initial testing was performed to optimize and determine various measurement parameters, and then the system was characterized using 252 Cf placed a various locations throughout the measurement system. Measurements were also performed with a 252 Cf source placed inside of small steel and HDPE shells to study the effect of moderation. These measurements compare favorably with their MCNPX model equivalent, making us confident that we can rely on the Monte Carlo simulation to predict the systematic uncertainty due to variations in response to material that may be localized at different points within a vessel.

  20. Short time effects of radiotherapy on lymphatic vessels and restorative lymphatic pathways: experimental approaches ina mouse model.

    Science.gov (United States)

    Pastouret, F; Lievens, P; Leduc, O; Bourgeois, P; Tournel, K; Lamote, J; Zirak, C; Leduc, A

    2014-06-01

    Radiotherapy (RT) is an important component in the therapeutic approach to oncologic conditions. This study presents the investigative results on the impact of RT on lymphatic vessels and on the regenerative response of the lymphatic system in a mouse model. We first irradiated 3 groups of ten mice using brachytherapy in a single treatment of 20 Gy. We then performed morphological examination of the irradiated lymphatic vessels using an in vivo microscopic transillumination technique at 2, 4, and 6 weeks. Next we evaluated lymphatic flow using lymphoscintigraphy and in vivo microscopy at 6 to 11 weeks in: 10 additional mice following irradiation as above (IR), in 10 mice following incision of a lymphatic vessel (I), and in a non-treated control group of 10 mice (N). Intact lymphatic vessels were observed in all mice at 2, 4, and 8 weeks following the single dose of radiotherapy in the first group of mice and normal lymphatic flow was fully restored in the irradiated (IR) and incised (I) mice indicating that the reparative substitution lymphatic pathways are functioning normally. We found that following irradiation with one dose of 20 Gy, lymphatic vessels were not visibly damaged and also that lymphatic flow was consistently restored and substitutive lymphatic pathways formed.

  1. Modeling Scala Media as a Pressure Vessel

    Science.gov (United States)

    Lepage, Eric; Olofsson, A.˚Ke

    2011-11-01

    The clinical condition known as endolymphatic hydrops is the swelling of scala media and may result in loss in hearing sensitivity consistent with other forms of low-frequency biasing. Because outer hair cells (OHCs) are displacement-sensitive and hearing levels tend to be preserved despite large changes in blood pressure and CSF pressure, it seems unlikely that the OHC respond passively to changes in static pressures in the chambers. This suggests the operation of a major feedback control loop which jointly regulates homeostasis and hearing sensitivity. Therefore the internal forces affecting the cochlear signal processing amplifier cannot be just motile responses. A complete account of the cochlear amplifier must include static pressures. To this end we have added a third, pressure vessel to our 1-D 140-segment, wave-digital filter active model of cochlear mechanics, incorporating the usual nonlinear forward transduction. In each segment the instantaneous pressure is the sum of acoustic pressure and global static pressure. The object of the model is to maintain stable OHC operating point despite any global rise in pressure in the third chamber. Such accumulated pressure is allowed to dissipate exponentially. In this first 3-chamber implementation we explore the possibility that acoustic pressures are rectified. The behavior of the model is critically dependent upon scaling factors and time-constants, yet by initial assumption, the pressure tends to accumulate in proportion to sound level. We further explore setting of the control parameters so that the accumulated pressure either stays within limits or may rise without bound.

  2. Development and application of surrogate model for assessment of ex-vessel debris bed dryout probability - 15157

    International Nuclear Information System (INIS)

    Yakush, S.E.; Lubchenko, N.T.; Kudinov, P.

    2015-01-01

    In this work we consider a water-cooled power reactor severe accident scenario with pressure vessel failure and subsequent release of molten corium. A surrogate model for prediction of dryout heat flux for ex-vessels debris beds of different shapes is developed. Functional form of dryout heat flux dependence on problem parameters is developed by the analysis of coolability problem in non-dimensional variables. It is shown that for a flat debris bed the dryout heat flux can be represented in terms of three 1-dimensional functions for which approximating formulas are found. For two-dimensional debris beds (cylindrical, conical, Gaussian heap, mound-shaped), an additional function taking into account the bed shape geometry is obtained from numerical simulations using DECOSIM code as a full model. With the surrogate model in hand, risk analysis of debris bed coolability is carried out by Monte Carlo sampling of the input parameters within selected ranges, with assumed distribution functions

  3. The New York Head-A precise standardized volume conductor model for EEG source localization and tES targeting.

    Science.gov (United States)

    Huang, Yu; Parra, Lucas C; Haufe, Stefan

    2016-10-15

    In source localization of electroencephalograpic (EEG) signals, as well as in targeted transcranial electric current stimulation (tES), a volume conductor model is required to describe the flow of electric currents in the head. Boundary element models (BEM) can be readily computed to represent major tissue compartments, but cannot encode detailed anatomical information within compartments. Finite element models (FEM) can capture more tissue types and intricate anatomical structures, but with the higher precision also comes the need for semi-automated segmentation, and a higher computational cost. In either case, adjusting to the individual human anatomy requires costly magnetic resonance imaging (MRI), and thus head modeling is often based on the anatomy of an 'arbitrary' individual (e.g. Colin27). Additionally, existing reference models for the human head often do not include the cerebro-spinal fluid (CSF), and their field of view excludes portions of the head and neck-two factors that demonstrably affect current-flow patterns. Here we present a highly detailed FEM, which we call ICBM-NY, or "New York Head". It is based on the ICBM152 anatomical template (a non-linear average of the MRI of 152 adult human brains) defined in MNI coordinates, for which we extended the field of view to the neck and performed a detailed segmentation of six tissue types (scalp, skull, CSF, gray matter, white matter, air cavities) at 0.5mm(3) resolution. The model was solved for 231 electrode locations. To evaluate its performance, additional FEMs and BEMs were constructed for four individual subjects. Each of the four individual FEMs (regarded as the 'ground truth') is compared to its BEM counterpart, the ICBM-NY, a BEM of the ICBM anatomy, an 'individualized' BEM of the ICBM anatomy warped to the individual head surface, and FEMs of the other individuals. Performance is measured in terms of EEG source localization and tES targeting errors. Results show that the ICBM-NY outperforms

  4. Numerical modeling of the pulse wave propagation in large blood vessels based on liquid and wall interaction

    International Nuclear Information System (INIS)

    Rup, K; Dróżdż, A

    2014-01-01

    The purpose of this article is to develop a non-linear, one-dimensional model of pulse wave propagation in the arterial cardiovascular system. The model includes partial differential equations resulting from the balance of mass and momentum for the fluid-filled area and the balance equation for the area of the wall and vessels. The considered mathematical model of pulse wave propagation in the thoracic aorta section takes into account the viscous dissipation of fluid energy, realistic values of parameters describing the physicochemical properties of blood and vessel wall. Boundary and initial conditions contain the appropriate information obtained from in vivo measurements. As a result of the numerical solution of the mass and momentum balance equations for the blood and the equilibrium equation for the arterial wall area, time- dependent deformation, respective velocity profiles and blood pressure were determined.

  5. Energy release and its containment within thin-walled, backed vessels

    International Nuclear Information System (INIS)

    Chambers, D.I.

    1983-01-01

    The problem adressed is the containment of a sudden release of energy of a magnitude up to 4 x 10 11 joules in a reusable vessel. The design process began by formulating dynamic models for both the input to such a vessel and the vessel itself and using these models to generate a general response. Modifications to the input and a more specific response are discussed. Computer codes used in calculations are described and listed

  6. Episodic vertigo resulting from vascular risk factors, cervical spondylosis and head rotation: Two case reports.

    Science.gov (United States)

    Owolabi, Mayowa O; Ogah, Okechukwu S; Ogunniyi, Adesola

    2007-01-01

    Vascular risk factors predispose to vertebrobasilar ischemia. Cervical osteophytes can impinge on the vertebral artery causing mechanical occlusion during head turning. Presentation with vertigo in such instances is a common finding. A patient with obesity, hyperlipidemia, hypertension, cervical spondylosis, and vertigo triggered by head rotation is presented. She responded to antihypertensive and lipid-lowering drugs, vestibular sedative and application of cervical collar. The second patient also exhibited similar features and responded to conservative treatment. Rotational vertebral artery occlusion resulting from cervical spondylosis in the presence of atherosclerosed collateral vessels is a cause of posterior circulation insufficiency manifesting as vertigo. The tetrad of vertigo resulting from vascular risk factors, cervical spondylosis, and head rotation is proposed for further research.

  7. Integrated Model of the Eye/Optic Nerve Head Biomechanical Environment

    Science.gov (United States)

    Ethier, C. R.; Feola, A.; Myers, J. G.; Nelson, E.; Raykin, J.; Samuels, B.

    2017-01-01

    Visual Impairment and Intracranial Pressure (VIIP) syndrome is a concern for long-duration space flight. Previously, it has been suggested that ocular changes observed in VIIP syndrome are related to the cephalad fluid shift that results in altered fluid pressures [1]. We are investigating the impact of changes in intracranial pressure (ICP) using a combination of numerical models, which simulate the effects of various environment conditions, including finite element (FE) models of the posterior eye. The specific interest is to understand how altered pressures due to gravitational changes affect the biomechanical environment of tissues of the posterior eye and optic nerve sheath. METHODS: Additional description of the numerical modeling is provided in the IWS abstract by Nelson et al. In brief, to simulate the effects of a cephalad fluid shift on the cardiovascular and ocular systems, we utilized a lumped-parameter compartment model of these systems. The outputs of this lumped-parameter model then inform boundary conditions (pressures) for a finite element model of the optic nerve head (Figure 1). As an example, we show here a simulation of postural change from supine to 15 degree head-down tilt (HDT), with primary outcomes being the predicted change in strains at the optic nerve head (ONH) region, specifically in the lamina cribrosa (LC), retrolaminar optic nerve, and prelaminar neural tissue (PLNT). The strain field can be decomposed into three orthogonal components, denoted as the first, second and third principal strains. We compare the peak tensile (first principal) and compressive (third principal) strains, since elevated strain alters cell phenotype and induces tissue remodeling. RESULTS AND CONCLUSIONS: Our lumped-parameter model predicted an IOP increase of c. 7 mmHg after 21 minutes of 15 degree HDT, which agreed with previous reports of IOP in HDT [1]. The corresponding FEM simulations predicted a relative increase in the magnitudes of the peak tensile

  8. Development of Reactor Vessel Bottom Mount Instrumentation Nozzle Routine Inspection Device

    Energy Technology Data Exchange (ETDEWEB)

    Khaled, Atya Ahmed Abdallah; Ihn, Namgung [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-10-15

    The primary coolant water of pressurized water reactors has created cracks in j-weld of penetrations with Alloy 600 through a process called primary water stress corrosion cracking. On October 6, 2013, BMI nozzle number 3 at Palo Verde Unit 3 (PVNGS-3) exhibited small white de-posits around the annulus. Nozzle attachment to the RV lower head is by J-groove weld to the inside penetration of the nozzle and the weld material is of Alloy 600 material. Above two cases clearly show the necessity of routine inspection of RV lower head penetration during refueling outage. Nondestructive inspection is generally performed to detect fine cracks or defects that may develop during operation. Defects usually occur at weld regions, hence most non-destructive inspection is to scan and check any defects or crack in the weld region. BMI nozzles at the bottom head of a nuclear reactor vessel (RV) are one of such area for inspection. But BMI nozzles have not been inspected during regular refuel outage due to the relative small size of BMI nozzle and limited impact of the consequences of BMI leak. However, there is growing concern since there have been leaks at nuclear power plants (NPPs) as well as recent operating experience. In this study, we propose a system that is conveniently used for nondestructive inspection of BMI nozzles during regular refueling outage without removing all the reactor internals. A 3D model of the inspection system was also developed along with the RV and internals which permits a virtual 3D simulation to check the design concept and usability of the system.

  9. Benchmark Calibration Tests Completed for Stirling Convertor Heater Head Life Assessment

    Science.gov (United States)

    Krause, David L.; Halford, Gary R.; Bowman, Randy R.

    2005-01-01

    A major phase of benchmark testing has been completed at the NASA Glenn Research Center (http://www.nasa.gov/glenn/), where a critical component of the Stirling Radioisotope Generator (SRG) is undergoing extensive experimentation to aid the development of an analytical life-prediction methodology. Two special-purpose test rigs subjected SRG heater-head pressure-vessel test articles to accelerated creep conditions, using the standard design temperatures to stay within the wall material s operating creep-response regime, but increasing wall stresses up to 7 times over the design point. This resulted in well-controlled "ballooning" of the heater-head hot end. The test plan was developed to provide critical input to analytical parameters in a reasonable period of time.

  10. Models for ductile crack initiation and tearing resistance under mode 1 loading in pressure vessel steels

    International Nuclear Information System (INIS)

    Jones, M.R.

    1988-06-01

    Micromechanistic models are presented which aim to predict plane strain ductile initiation toughness, tearing resistance and notched bar fracture strains in pressure vessel steels under monotonically increasing tensile (mode 1) loading. The models for initiation toughness and tearing resistance recognize that ductile fracture proceeds by the growth and linkage of voids with the crack-tip. The models are shown to predict the trend of initiation toughness with inclusion spacing/size ratio and can bound the available experimental data. The model for crack growth can reproduce the tearing resistance of a pressure vessel steel up to and just beyond crack growth initiation. The fracture strains of notched bars pulled in tension are shown to correspond to the achievement of a critical volume fraction of voids. This criterion is combined with the true stress - true strain history of a material point ahead of a blunting crack-tip to predict the initiation toughness. An attempt was made to predict the fracture strains of notched tensile bars by adopting a model which predicts the onset of a shear localization phenomenon. Fracture strains of the correct order are computed only if a ''secondary'' void nucleation event at carbide precipitates is taken into account. (author)

  11. Experimental Creep Life Assessment for the Advanced Stirling Convertor Heater Head

    Science.gov (United States)

    Krause, David L.; Kalluri, Sreeramesh; Shah, Ashwin R.; Korovaichuk, Igor

    2010-01-01

    The United States Department of Energy is planning to develop the Advanced Stirling Radioisotope Generator (ASRG) for the National Aeronautics and Space Administration (NASA) for potential use on future space missions. The ASRG provides substantial efficiency and specific power improvements over radioisotope power systems of heritage designs. The ASRG would use General Purpose Heat Source modules as energy sources and the free-piston Advanced Stirling Convertor (ASC) to convert heat into electrical energy. Lockheed Martin Corporation of Valley Forge, Pennsylvania, is integrating the ASRG systems, and Sunpower, Inc., of Athens, Ohio, is designing and building the ASC. NASA Glenn Research Center of Cleveland, Ohio, manages the Sunpower contract and provides technology development in several areas for the ASC. One area is reliability assessment for the ASC heater head, a critical pressure vessel within which heat is converted into mechanical oscillation of a displacer piston. For high system efficiency, the ASC heater head operates at very high temperature (850 C) and therefore is fabricated from an advanced heat-resistant nickel-based superalloy Microcast MarM-247. Since use of MarM-247 in a thin-walled pressure vessel is atypical, much effort is required to assure that the system will operate reliably for its design life of 17 years. One life-limiting structural response for this application is creep; creep deformation is the accumulation of time-dependent inelastic strain under sustained loading over time. If allowed to progress, the deformation eventually results in creep rupture. Since creep material properties are not available in the open literature, a detailed creep life assessment of the ASC heater head effort is underway. This paper presents an overview of that creep life assessment approach, including the reliability-based creep criteria developed from coupon testing, and the associated heater head deterministic and probabilistic analyses. The approach also

  12. Assessment of motion effects on the FPSO (Floating, Production, Storage and Offloading) vessel Terra Nova

    Energy Technology Data Exchange (ETDEWEB)

    Cheung, B.; Hofer, K. [Defence Research and Development Canada, Toronto, ON (Canada); Brooks, C.J. [Survival Systems Group Ltd., Dartmouth, NS (Canada)

    2002-10-01

    A study was conducted to define the incidence and severity of seasickness, motion-induced fatigue and task performance problems encountered on the Floating, Production, Storage, Offshore (FPSO) vessel which Petro-Canada operates in the Grands Banks of Newfoundland at the Terra Nova Field. The FPSO vessel is tethered to the oil well head by flexible couplings and is subjected to severe wave motion at sea. Crew members living and working aboard the FPSO vessel are exposed to more severe weather motion compared to those on fixed installation platforms, particularly during the winter months. The study involved a questionnaire to determine if seasickness is a problem and whether specific ship motions affect sleep, mental and physical performance on the vessel. Ship motion data was obtained through sensors mounted on the bow of the vessel. Respondents revealed that the incidence and severity of motion sickness and sleep disturbance ranged from slight to moderate. The correlation between sleep disturbance and ship motion was high. Problems in task performance ranged from loss of concentration, decision making and memory disorders and task completion problems. The number of safety, health and performance issues increased with bad weather conditions. One of the objectives of this study is to develop recommendations to provide operations guidance to improve comfort and performance on FPSO vessels. 13 tabs., 7 figs.

  13. The method to make the three dimensional anatomical pattern of hepatic vessels by stereo angiography

    International Nuclear Information System (INIS)

    Mutou, Haruomi; Kobayashi, Seiichiro; Yamada, Akiyoshi; Takasaki, Takeshi; Isobe, Yoshinori; Tanaka, Seiichi; Saeki, Shin; Yoshida, Masanori

    1986-01-01

    For the Past few years, there has been a big advance in the hepatic surgery. Now, small resection, such as segmentectomy or subsegmentectomy, is performed routinely. Based on this tendency, hepatic surgeons request more details, more stereographic findings of hepatic vessels to heaptic angiography. Especially three dimensional combined anatomical pattern of the hepatic artery, portal vein and hepatic vein is strongly needed. We have tried three dimensional computer graphic of hepatic vessels since few years ago, using the personal computer, digitizer with clear screen, commercially available 3D software and my own program. We use three groups of angiographic films, that is the hepatic artery, portal vein and hepatic vein with IVC, which were taken by stereoangiography. The depth of each poits of vessels are calculated by the way described in Fig 3. Using these points, the 3D software, '3DPROGATS', can make the anatomical pattern of combined hepatic vessels on TV display. And then we can also perform rotation, heading, bank, zooming, hidden line elimination freely for this picture. Out of necessity as hepatic surgeons, we make a simple system for 3D computer graphic of heptic vessels. At present, the image is somewhat rough, but clinically it is relatively effective. In this report we want to explain our method and to show the anatomical pattern of hepatic vessels of case of hepatoma. (author)

  14. Vascular Patterns in Iguanas and Other Squamates: Blood Vessels and Sites of Thermal Exchange.

    Directory of Open Access Journals (Sweden)

    William Ruger Porter

    Full Text Available Squamates use the circulatory system to regulate body and head temperatures during both heating and cooling. The flexibility of this system, which possibly exceeds that of endotherms, offers a number of physiological mechanisms to gain or retain heat (e.g., increase peripheral blood flow and heart rate, cooling the head to prolong basking time for the body as well as to shed heat (modulate peripheral blood flow, expose sites of thermal exchange. Squamates also have the ability to establish and maintain the same head-to-body temperature differential that birds, crocodilians, and mammals demonstrate, but without a discrete rete or other vascular physiological device. Squamates offer important anatomical and phylogenetic evidence for the inference of the blood vessels of dinosaurs and other extinct archosaurs in that they shed light on the basal diapsid condition. Given this basal positioning, squamates likewise inform and constrain the range of physiological thermoregulatory mechanisms that may have been found in Dinosauria. Unfortunately, the literature on squamate vascular anatomy is limited. Cephalic vascular anatomy of green iguanas (Iguana iguana was investigated using a differential-contrast, dual-vascular injection (DCDVI technique and high-resolution X-ray microcomputed tomography (μCT. Blood vessels were digitally segmented to create a surface representation of vascular pathways. Known sites of thermal exchange, consisting of the oral, nasal, and orbital regions, were given special attention due to their role in brain and cephalic thermoregulation. Blood vessels to and from sites of thermal exchange were investigated to detect conserved vascular patterns and to assess their ability to deliver cooled blood to the dural venous sinuses. Arteries within sites of thermal exchange were found to deliver blood directly and through collateral pathways. The venous drainage was found to have multiple pathways that could influence neurosensory

  15. Vascular Patterns in Iguanas and Other Squamates: Blood Vessels and Sites of Thermal Exchange.

    Science.gov (United States)

    Porter, William Ruger; Witmer, Lawrence M

    2015-01-01

    Squamates use the circulatory system to regulate body and head temperatures during both heating and cooling. The flexibility of this system, which possibly exceeds that of endotherms, offers a number of physiological mechanisms to gain or retain heat (e.g., increase peripheral blood flow and heart rate, cooling the head to prolong basking time for the body) as well as to shed heat (modulate peripheral blood flow, expose sites of thermal exchange). Squamates also have the ability to establish and maintain the same head-to-body temperature differential that birds, crocodilians, and mammals demonstrate, but without a discrete rete or other vascular physiological device. Squamates offer important anatomical and phylogenetic evidence for the inference of the blood vessels of dinosaurs and other extinct archosaurs in that they shed light on the basal diapsid condition. Given this basal positioning, squamates likewise inform and constrain the range of physiological thermoregulatory mechanisms that may have been found in Dinosauria. Unfortunately, the literature on squamate vascular anatomy is limited. Cephalic vascular anatomy of green iguanas (Iguana iguana) was investigated using a differential-contrast, dual-vascular injection (DCDVI) technique and high-resolution X-ray microcomputed tomography (μCT). Blood vessels were digitally segmented to create a surface representation of vascular pathways. Known sites of thermal exchange, consisting of the oral, nasal, and orbital regions, were given special attention due to their role in brain and cephalic thermoregulation. Blood vessels to and from sites of thermal exchange were investigated to detect conserved vascular patterns and to assess their ability to deliver cooled blood to the dural venous sinuses. Arteries within sites of thermal exchange were found to deliver blood directly and through collateral pathways. The venous drainage was found to have multiple pathways that could influence neurosensory tissue temperature

  16. Robot-Assisted Free Flap in Head and Neck Reconstruction

    Directory of Open Access Journals (Sweden)

    Han Gyeol Song

    2013-07-01

    Full Text Available Background  Robots have allowed head and neck surgeons to extirpate oropharyngealtumors safely without the need for lip-split incision or mandibulotomy. Using robots inoropharyngealreconstruction is newbut essentialfor oropharyngeal defectsthatresultfromrobotic tumor excision. We report our experience with robotic free-flap reconstruction ofhead and neck defectsto exemplify the necessity forrobotic reconstruction.Methods  We investigated head and neck cancer patients who underwent ablation surgeryand free-flap reconstruction by robot. Between July 1, 2011 andMarch 31, 2012, 5 caseswereperformed and patient demographics, location of tumor, pathologic stage, reconstructionmethods, flap size, recipient vessel, necessary pedicle length, and operation time wereinvestigated.Results  Among five free-flap reconstructions, four were radial forearm free flaps and onewas an anterolateral thigh free-flap. Four flaps used the superior thyroid artery and oneflap used a facial artery as the recipient vessel. The average pedicle length was 8.8 cm. Flapinsetting and microanastomosis were achieved using a specially manufactured roboticinstrument. The total operation timewas 1,041.0 minutes(range, 814 to 1,132 minutes, andcomplicationsincluding flap necrosis, hematoma, andwound dehiscence did not occur.Conclusions  Thisstudy demonstratesthe clinically applicable use ofrobotsin oropharyngealreconstruction, especially using a free flap. A robot can assist the operator in insettingthe flap at a deep portion of the oropharynx without the need to perform a traditionalmandibulotomy. Robot-assisted reconstruction may substitute for existing surgical methodsand is accepted asthemost up-to-datemethod.

  17. Design and Computational Fluid Dynamics Optimization of the Tube End Effector for Reactor Pressure Vessel Head Type VVER-1000

    International Nuclear Information System (INIS)

    Novosel, D.

    2006-01-01

    In this paper is presented development and optimization of the tube end effector design which should consist of 4 ultrasonic transducers, 4 Eddy Current's transducers and Radiation Proof Dot Camera. Basically, designing was conducted by main input requests, such as: inner diameter of a tested reactor pressure vessel head penetration tube, dimensions of a transducers and maximum allowable vertical movement of a manipulator connection rod in order to cover all inner tube surface. As is obvious, for ultrasonic testing should be provided the thin layer of liquid material (in our case water was chosen) which is necessary to make physical contact between transducer surface and investigated inner tube surface. By help of Computational Fluid Dynamics, determined were parameters of geometry, as the most important factor of transducer housing, hydraulically parameters for water supply and primary drain together implemented into this housing, movement of the end effectors (vertical and cylindrical) and finally, necessary equipment which has to provide all hydraulically and pneumatic requirements. As the cylindrical surface of the inner tube diameter was liquefied and contact between transducer housing and tested tube wasn't ideally covered, water leakage could occur in downstream direction. To reduce water leakage, which is highly contaminated, developed was second water drain by diffuser assembly which is driven by Venturi pipe, commercially called vacuum generator. Using the Computational Fluid Dynamic, obtained was optimized geometry of diffuser control volume with the highest efficiency, in other words, unobstructed fluid flux. Afterwards, the end effectors system was synchronized to the existing operable system for NDT methods all invented and designed by INETEC. (author)

  18. Dual-energy CT head bone and hard plaque removal for quantification of calcified carotid stenosis: utility and comparison with digital subtraction angiography

    International Nuclear Information System (INIS)

    Uotani, Kensuke; Watanabe, Yoshiyuki; Higashi, Masahiro; Nakazawa, Tetsuro; Kono, Atsushi K.; Hori, Yoshiro; Fukuda, Tetsuya; Kanzaki, Suzu; Yamada, Naoaki; Naito, Hiroaki; Itoh, Toshihide; Sugimura, Kazuro

    2009-01-01

    We evaluated quantification of calcified carotid stenosis by dual-energy (DE) CTA and dual-energy head bone and hard plaque removal (DE hard plaque removal) and compared the results to those of digital subtraction angiography (DSA). Eighteen vessels (13 patients) with densely calcified carotid stenosis were examined by dual-source CT in the dual-energy mode (tube voltages 140 kV and 80 kV). Head bone and hard plaques were removed from the dual-energy images by using commercial software. Carotid stenosis was quantified according to NASCET criteria on MIP images and DSA images at the same plane. Correlation between DE CTA and DSA was determined by cross tabulation. Accuracies for stenosis detection and grading were calculated. Stenosis could be evaluated in all vessels by DE CTA after applying DE hard plaque removal. In contrast, conventional CTA failed to show stenosis in 13 out of 18 vessels due to overlapping hard plaque. Good correlation between DE plaque removal images and DSA images was observed (r 2 =0.9504) for stenosis grading. Sensitivity and specificity to detect hemodynamically relevant (>70%) stenosis was 100% and 92%, respectively. Dual-energy head bone and hard plaque removal is a promising tool for the evaluation of densely calcified carotid stenosis. (orig.)

  19. Development of a child head analytical dynamic model considering cranial nonuniform thickness and curvature - Applying to children aged 0-1 years old.

    Science.gov (United States)

    Li, Zhigang; Ji, Cheng; Wang, Lishu

    2018-07-01

    Although analytical models have been used to quickly predict head response under impact condition, the existing models generally took the head as regular shell with uniform thickness which cannot account for the actual head geometry with varied cranial thickness and curvature at different locations. The objective of this study is to develop and validate an analytical model incorporating actual cranial thickness and curvature for child aged 0-1YO and investigate their effects on child head dynamic responses at different head locations. To develop the new analytical model, the child head was simplified into an irregular fluid-filled shell with non-uniform thickness and the cranial thickness and curvature at different locations were automatically obtained from CT scans using a procedure developed in this study. The implicit equation of maximum impact force was derived as a function of elastic modulus, thickness and radius of curvature of cranium. The proposed analytical model are compared with cadaver test data of children aged 0-1 years old and it is shown to be accurate in predicting head injury metrics. According to this model, obvious difference in injury metrics were observed among subjects with the same age, but different cranial thickness and curvature; and the injury metrics at forehead location are significant higher than those at other locations due to large thickness it owns. The proposed model shows good biofidelity and can be used in quickly predicting the dynamics response at any location of head for child younger than 1 YO. Copyright © 2018 Elsevier B.V. All rights reserved.

  20. Analytical modelling of soccer heading

    Indian Academy of Sciences (India)

    ... game is that the players are permitted to use their head to direct the ball during ... method in assessing the cognitive functions that can be applied not only to ... It is attached to a spring (stiffness, k1) and a dashpot (damping coefficient, c1).

  1. A Statistical Model of Head Asymmetry in Infants with Deformational Plagiocephaly

    DEFF Research Database (Denmark)

    Lanche, Stéphanie; Darvann, Tron Andre; Ólafsdóttir, Hildur

    2007-01-01

    Deformational plagiocephaly is a term describing cranial asymmetry and deformation commonly seen in infants. The purpose of this work was to develop a methodology for assessment and modelling of head asymmetry. The clinical population consisted of 38 infants for whom 3-dimensional surface scans...... quantitative description of the asymmetry present in the dataset....

  2. Histopathological changes in the head kidney induced by cadmium in a neotropical fish Colossoma macropomum

    Directory of Open Access Journals (Sweden)

    R. Salazar-Lugo

    2013-12-01

    Full Text Available We evaluated the effect of cadmium (Cd on the structure and function of the head kidney in the freshwater fish Colossoma macropomum (C. macropomum. Juveniles were exposed to 0.1 mg/L CdCl2 for 31 days. Blood samples were examined using hematological tests and head kidney histology was determined by light microscopy. The concentration of Cd in the head and trunk kidneys was measured using an atomic absorption spectrophotometer. Cd produced histopathological changes in the head kidney, the most evident of these being: the thickening of the vein wall, an increase in the number of basophils/mast cells close to blood vessels and a severe depletion of hematopoietic precursors especially the granulopoietic series. In the blood, a decrease in the total leucocytes and hemoglobin concentration was observed. Cd-exposed fish showed higher Cd concentrations in the trunk kidney than the head kidney. In conclusion, exposure to Cd affected precursor hematopoietic cells in C. macropomum.

  3. Modeling the Role of the Glymphatic Pathway and Cerebral Blood Vessel Properties in Alzheimer’s Disease Pathogenesis

    Science.gov (United States)

    Kyrtsos, Christina Rose; Baras, John S.

    2015-01-01

    Alzheimer’s disease (AD) is the most common cause of dementia in the elderly, affecting over 10% population over the age of 65 years. Clinically, AD is described by the symptom set of short term memory loss and cognitive decline, changes in mentation and behavior, and eventually long-term memory deficit as the disease progresses. On imaging studies, significant atrophy with subsequent increase in ventricular volume have been observed. Pathology on post-mortem brain specimens demonstrates the classic findings of increased beta amyloid (Aβ) deposition and the presence of neurofibrillary tangles (NFTs) within affected neurons. Neuroinflammation, dysregulation of blood-brain barrier transport and clearance, deposition of Aβ in cerebral blood vessels, vascular risk factors such as atherosclerosis and diabetes, and the presence of the apolipoprotein E4 allele have all been identified as playing possible roles in AD pathogenesis. Recent research has demonstrated the importance of the glymphatic system in the clearance of Aβ from the brain via the perivascular space surrounding cerebral blood vessels. Given the variety of hypotheses that have been proposed for AD pathogenesis, an interconnected, multilayer model offers a unique opportunity to combine these ideas into a single unifying model. Results of this model demonstrate the importance of vessel stiffness and heart rate in maintaining adequate clearance of Aβ from the brain. PMID:26448331

  4. Vessel-guided airway tree segmentation

    DEFF Research Database (Denmark)

    Lo, Pechin Chien Pau; Sporring, Jon; Ashraf, Haseem

    2010-01-01

    This paper presents a method for airway tree segmentation that uses a combination of a trained airway appearance model, vessel and airway orientation information, and region growing. We propose a voxel classification approach for the appearance model, which uses a classifier that is trained to di...

  5. VISA-2, Reactor Vessel Failure Probability Under Thermal Shock

    International Nuclear Information System (INIS)

    Simonen, F.; Johnson, K.

    1992-01-01

    1 - Description of program or function: VISA2 (Vessel Integrity Simulation Analysis) was developed to estimate the failure probability of nuclear reactor pressure vessels under pressurized thermal shock conditions. The deterministic portion of the code performs heat transfer, stress, and fracture mechanics calculations for a vessel subjected to a user-specified temperature and pressure transient. The probabilistic analysis performs a Monte Carlo simulation to estimate the probability of vessel failure. Parameters such as initial crack size and position, copper and nickel content, fluence, and the fracture toughness values for crack initiation and arrest are treated as random variables. Linear elastic fracture mechanics methods are used to model crack initiation and growth. This includes cladding effects in the heat transfer, stress, and fracture mechanics calculations. The simulation procedure treats an entire vessel and recognizes that more than one flaw can exist in a given vessel. The flaw model allows random positioning of the flaw within the vessel wall thickness, and the user can specify either flaw length or length-to-depth aspect ratio for crack initiation and arrest predictions. The flaw size distribution can be adjust on the basis of different inservice inspection techniques and inspection conditions. The toughness simulation model includes a menu of alternative equations for predicting the shift in the reference temperature of the nil-ductility transition. 2 - Method of solution: The solution method uses closed form equations for temperatures, stresses, and stress intensity factors. A polynomial fitting procedure approximates the specified pressure and temperature transient. Failure probabilities are calculated by a Monte Carlo simulation. 3 - Restrictions on the complexity of the problem: Maxima of 30 welds. VISA2 models only the belt-line (cylindrical) region of a reactor vessel. The stresses are a function of the radial (through-wall) coordinate only

  6. Clinical usefulness of a newly-developed head and neck surface coil for MR imaging

    International Nuclear Information System (INIS)

    Shimada, Morio; Kogure, Takashi; Hayashi, Sanshin

    1995-01-01

    To obtain correct diagnosis at early stages of cervical lymph node swelling, especially cases with suspected epipharyngeal carcinoma, and cerebral arterial sclerotic diseases, high-quality MR images visualizing the entire head and neck structures and vessels are of crucial importance. When obtaining images of head and neck regions using a head coil, signal intensity (SI) and signal to noise ratio (SNR) of regions below the hypopharynx are weakened. Moreover, when obtaining images of head and neck regions using an anterior neck coil, SI and SNR of upper regions of epipharynx are also weakened. In an attempt to solve these problems, we developed a new head and neck surface coil for MR imaging. With this new coil we were able to obtain better images (153 cases) from regions below the hypopharynx to the upper regions of the epipharynx in the same time as images obtained using the head coil and anterior neck coil. 2D TOF MR angiographic images (11 cases) obtained by the head and neck surface coil are superior to 2D TOF angiographic images obtained by the anterior neck coil. MR images obtained with this improved method are valuable in the evaluation and management of head and neck region disease. (author)

  7. Hypercholesterolemia induced cerebral small vessel disease.

    Science.gov (United States)

    Kraft, Peter; Schuhmann, Michael K; Garz, Cornelia; Jandke, Solveig; Urlaub, Daniela; Mencl, Stine; Zernecke, Alma; Heinze, Hans-Jochen; Carare, Roxana O; Kleinschnitz, Christoph; Schreiber, Stefanie

    2017-01-01

    While hypercholesterolemia plays a causative role for the development of ischemic stroke in large vessels, its significance for cerebral small vessel disease (CSVD) remains unclear. We thus aimed to understand the detailed relationship between hypercholesterolemia and CSVD using the well described Ldlr-/- mouse model. We used Ldlr-/- mice (n = 16) and wild-type (WT) mice (n = 15) at the age of 6 and 12 months. Ldlr-/- mice develop high plasma cholesterol levels following a high fat diet. We analyzed cerebral capillaries and arterioles for intravascular erythrocyte accumulations, thrombotic vessel occlusions, blood-brain barrier (BBB) dysfunction and microbleeds. We found a significant increase in the number of erythrocyte stases in 6 months old Ldlr-/- mice compared to all other groups (P hypercholesterolemia is related to a thrombotic CSVD phenotype, which is different from hypertension-related CSVD that associates with a hemorrhagic CSVD phenotype. Our data demonstrate a relationship between hypercholesterolemia and the development of CSVD. Ldlr-/- mice appear to be an adequate animal model for research into CSVD.

  8. Scatter modelling of fracture toughness data for reactor pressure vessel structural integrity assessment

    International Nuclear Information System (INIS)

    Pesoz, M.

    1997-01-01

    In the last decade, there has been an increasing interest at EDF in developing and applying probabilistic methods for a variety of purposes. In the field of structural integrity and reliability they are used to evaluate the effect of deterioration due to ageing mechanisms, mainly on major passive structural components such as reactor pressure vessel, steam generator and piping in nuclear plants. Such approaches provide an attractive supplement to the more conventional deterministic method, based upon pessimistic assumptions, that give results too far from reality to support effective decisions. In addition to deterministic calculations, a Probabilistic Fracture Mechanics model has been developed in order to analyse the risk of brittle failure of the reactor pressure vessel and to perform sensitivity studies. The material fracture toughness (K IC ) uncertainty appears to be strongly influencing the probability of failure under accidental conditions. Up to now, this parameter is determined from the RCC-M code reference curve, which is the same as the ASME reference curve. But an important issue when performing probabilistic analysis is the correct statistical modelling of input parameters. That's why modelling works have been carried out using results of fracture toughness tests performed for demonstrating the validity of the reference curve. This paper presents the statistical treatments that have been performed to model the scatter of temperature dependent parameter (K IC (T). A specific data base containing a few hundreds of French and US results have been carried and Weibull models have been fitted, based on various master curve equations (K. Wallin (Senior Adviser at the Technical Research Centre of Finland) or RCC-M types). (author)

  9. 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

  10. Numerical studies of large penetrations and closures for containment vessels subjected to loadings beyond the design basis

    International Nuclear Information System (INIS)

    Kulak, R.F.; Hsieh, B.J.; Kennedy, J.M.; Ash, J.E.; McLennan, G.A.

    1984-01-01

    Numerical simulations of the macro-deformations of the sealing surfaces (gasketed junctures) of a PWR steel containment vessel's equipment hatch and a BWR Mk II containment vessel head have been performed. Results for the equipment hatch juncture indicate that the rotations of the hatch cover and penetration sleeve must be accounted for when performing leakage analysis because they can effect the compression of the gasket even though the gasket is in a pressure-seated configuration. Results from a leakage analysis indicated that excessive leakage can occur if the surface roughness is high and/or the compression set is high. Results for the Mk II head show that both the temperature and pressure loadings must be taken into account to obtain realistic responses. The temperature difference between the flanges and bolts has the important net effect of keeping the gasketed juncture closed, that is in metal-to-metal contact. Due to the high accident temperature, the gasket itself was found to achieve 100% compression set and thus could not perform its sealing function within the juncture

  11. High-resolution 3D Magnetic Resonance angiography in the evaluation of neck vessels and intracranial circulation

    International Nuclear Information System (INIS)

    Villa, A.; Di Guglielmo, L.; Campani, R.; Nicolato, A.; D'Amato, M.; Rodriguez y Balena, R.

    1991-01-01

    Magnetic Resonance Angiography (MRA) is a modern vascular imaging technique which allows the non-invasive and direct imaging of vessels. The authors aimed at evaluating the diagnostic accuracy of MRA in the study of pathologic conditions in the neck and intracranial vessels; spatial resolution of the technique was also investigated. Twenty-four healthy volunteers and 82 patients suffering from various diseases of the head and neck vessels were included in the study. First of all, MRA capabilities ware investigated in visualizing normal vessels of both neck and intracranial circle. The diagnostic accuracy of the method was then evaluated in the study of vascular diseases, and the results compared with conventional/digital angiographic findings. The comparison demonstrated how stenoses and atherosclerotic plaques tend to be overestimated by MRA because of technical artifacts inherent to the technique itself, whereas vascular ulcerations and aneurysms are frequently underestimated. However, this data was steady and therefore evaluable- the exact knowledge of the artifacts making diagnosis reliable. The diagnostic and technical problems relative to the various vascular diseases are discussed. Finally, several hypotheses of diagnostic iter are suggested

  12. Study of the influence of the laterality of mobile phone use on the SAR induced in two head models

    Science.gov (United States)

    Ghanmi, Amal; Varsier, Nadège; Hadjem, Abdelhamid; Conil, Emmanuelle; Picon, Odile; Wiart, Joe

    2013-05-01

    The objective of this paper is to investigate and to analyse the influence of the laterality of mobile phone use on the exposure of the brain to radio-frequencies (RF) and electromagnetic fields (EMF) from different mobile phone models using the finite-difference time-domain (FDTD) method. The study focuses on the comparison of the specific absorption rate (SAR) induced on the right and left sides of two numerical adult and child head models. The heads are exposed by both phone models operating in GSM frequency bands for both ipsilateral and contralateral configurations. A slight SAR difference between the two sides of the heads is noted. The results show that the variation between the left and the right sides is more important at 1800 MHz for an ipsilateral use. Indeed, at this frequency, the variation can even reach 20% for the SAR10g and the SAR1g induced in the head and in the brain, respectively. Moreover, the average SAR induced by the mobile phone in the half hemisphere of the brain in ipsilateral exposure is higher than in contralateral exposure. Owing to the superficial character of energy deposition at 1800 MHz, this difference in the SAR induced for the ipsilateral and contralateral usages is more significant at 1800 MHz than at 900 MHz. The results have shown that depending on the phantom head models, the SAR distribution in the brain can vary because of differences in anatomical proportions and in the geometry of the head models. The induced SAR in child head and in sub-regions of the brain is significantly higher (up to 30%) compared to the adult head. This paper confirms also that the shape/design of the mobile and the location of the antenna can have a large influence at high frequency on the exposure of the brain, particularly on the SAR distribution and on the distinguished brain regions.

  13. Demarcation of inland vessels' limit off Mormugao port region, India: A pilot study for the safety of inland vessels using wave modelling

    Digital Repository Service at National Institute of Oceanography (India)

    Vethamony, P.; Aboobacker, V.M.; Sudheesh, K.; Babu, M.T.; AshokKumar, A.

    The Ministry of Shipping desires to revise the inland vessels' limit (IVL) notification based on scientific rationale to improve the safety of vessels and onboard personnel. The Mormugao port region extending up to the Panaji was considered...

  14. Effect of head restraint backset on head-neck kinematics in whiplash.

    Science.gov (United States)

    Stemper, Brian D; Yoganandan, Narayan; Pintar, Frank A

    2006-03-01

    Although head restraints were introduced in the 1960s as a countermeasure for whiplash, their limited effectiveness has been attributed to incorrect positioning. The effect of backset on cervical segmental angulations, which were previously correlated with spinal injury, has not been delineated. Therefore, the practical restraint position to minimize injury remains unclear. A parametric study of increasing head restraint backset between 0 and 140mm was conducted using a comprehensively validated computational model. Head retraction values increased with increasing backset, reaching a maximum value of 53.5mm for backsets greater than 60mm. Segmental angulation magnitudes, greatest at levels C5-C6 and C6-C7, reached maximum values during the retraction phase and increased with increasing backset. Results were compared to a previously published head restraint rating system, wherein lower cervical extension magnitudes from this study exceeded mean physiologic limits for restraint positions rated good, acceptable, marginal, and poor. As head restraint contact was the limiting factor in head retraction and segmental angulations, the present study indicates that minimizing whiplash injury may be accomplished by limiting head restraint backset to less than 60mm either passively or actively after impact.

  15. Meteor head echoes - observations and models

    Directory of Open Access Journals (Sweden)

    A. Pellinen-Wannberg

    2005-01-01

    Full Text Available Meteor head echoes - instantaneous echoes moving with the velocities of the meteors - have been recorded since 1947. Despite many attempts, this phenomenon did not receive a comprehensive theory for over 4 decades. The High Power and Large Aperture (HPLA features, combined with present signal processing and data storage capabilities of incoherent scatter radars, may give an explanation for the old riddle. The meteoroid passage through the radar beam can be followed with simultaneous spatial-time resolution of about 100m-ms class. The current views of the meteor head echo process will be presented and discussed. These will be related to various EISCAT observations, such as dual-frequency target sizes, altitude distributions and vector velocities.

  16. Fluid and mass transport in a single lymphatic blood vessel

    International Nuclear Information System (INIS)

    Bestman, A.R.

    1987-08-01

    The problem considers the single blood vessel model in pulmonary circulation in the presence of gravitation and mass transfer. The tissue surrounding the blood vessel is modelled as a permeable medium distinct from the blood vessel which is a normal free space. On the assumption that the mass concentration varies slowly at the interface between the blood vessel and the tissue, the problem is tackled by asymptotic approximation. A crucial point of the analysis is the dependence of the flow variables on the permeability K of the tissue in a completely arbitrary manner. A primary conjecture of the study is the intimacy of the pathological pulmonary edema and the parameter K. (author). 4 refs

  17. 33 CFR 90.3 - Pushing vessel and vessel being pushed: Composite unit.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 1 2010-07-01 2010-07-01 false Pushing vessel and vessel being... HOMELAND SECURITY INLAND NAVIGATION RULES INLAND RULES: INTERPRETATIVE RULES § 90.3 Pushing vessel and vessel being pushed: Composite unit. Rule 24(b) of the Inland Rules states that when a pushing vessel and...

  18. 33 CFR 82.3 - Pushing vessel and vessel being pushed: Composite unit.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 1 2010-07-01 2010-07-01 false Pushing vessel and vessel being... HOMELAND SECURITY INTERNATIONAL NAVIGATION RULES 72 COLREGS: INTERPRETATIVE RULES § 82.3 Pushing vessel and vessel being pushed: Composite unit. Rule 24(b) of the 72 COLREGS states that when a pushing vessel and a...

  19. Confinement Vessel Assay System: Calibration and Certification Report

    Energy Technology Data Exchange (ETDEWEB)

    Frame, Katherine C. [Los Alamos National Laboratory; Bourne, Mark M. [Los Alamos National Laboratory; Crooks, William J. [Los Alamos National Laboratory; Evans, Louise [Los Alamos National Laboratory; Gomez, Cipriano [Retired CMR-OPS: OPERATIONS; Mayo, Douglas R. [Los Alamos National Laboratory; Miko, David K. [Los Alamos National Laboratory; Salazar, William R. [Los Alamos National Laboratory; Stange, Sy [Los Alamos National Laboratory; Vigil, Georgiana M. [Los Alamos National Laboratory

    2012-07-17

    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 to 2 inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of 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 {sup 239}Pu equivalent in a vessel for safeguards termination. The system was calibrated in three different mass regions (low, medium, and high) to cover the entire plutonium mass range that will be assayed. The low mass calibration and medium mass calibration were verified for material positioned in the center of an empty vessel. The systematic uncertainty due to position bias was estimated using an MCNPX model to simulate the response of the system to material localized at various points along the inner surface of the vessel. The background component due to cosmic ray spallation was determined by performing measurements of an empty vessel and comparing to measurements in the same location with no vessel present. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements of CVs before and after cleanout.

  20. Confinement Vessel Assay System: Calibration and Certification Report

    International Nuclear Information System (INIS)

    Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Gomez, Cipriano; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; 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 to 2 inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of 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 system was calibrated in three different mass regions (low, medium, and high) to cover the entire plutonium mass range that will be assayed. The low mass calibration and medium mass calibration were verified for material positioned in the center of an empty vessel. The systematic uncertainty due to position bias was estimated using an MCNPX model to simulate the response of the system to material localized at various points along the inner surface of the vessel. The background component due to cosmic ray spallation was determined by performing measurements of an empty vessel and comparing to measurements in the same location with no vessel present. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements of CVs before and after cleanout.