Method of burying vessel containing radioactive waste

International Nuclear Information System (INIS)

Koga, Yoshihito.

1989-01-01

A float having an inert gas sealed therein is attached to a tightly closed vessel containing radioactive wastes. The vessel is inserted and kept in a small hole for burying the tightly closed vessel in an excavated shaft in rocks such as of granite or rock salts, while filling bentonite as shielding material therearound. In this case, the float is so adjusted that the apparent specific gravity is made equal or nearer between the tightly closed vessel and the bentonite, so that the rightly closed vessel does not sink and cause direct contact with the rocks even if bentonite flows due to earthquakes, etc. This can prevent radioactivity contamination through water in the rocks. (S.K.)

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.)

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

Closed-flow column experiments—Insights into solute transport provided by a damped oscillating breakthrough behavior

Science.gov (United States)

Ritschel, Thomas; Totsche, Kai Uwe

2016-03-01

Transport studies that employ column experiments in closed-flow mode complement classical approaches by providing new characteristic features observed in the solute breakthrough and equilibrium between liquid and solid phase. Specific to the closed-flow mode is the recirculation of the effluent to the inflow via a mixing vessel. Depending on the ratio of volumes of mixing vessel and water-filled pore space, a damped oscillating solute concentration emerges in the effluent and mixing vessel. The oscillation characteristics, e.g., frequency, amplitude, and damping, allow for the investigation of solute transport in a similar fashion as known for classical open-flow column experiments. However, the closed loop conserves substances released during transport within the system. In this way, solute and porous medium can equilibrate with respect to physicochemical conditions. With this paper, the features emerging in the breakthrough curves of saturated column experiments run in closed-flow mode and methods of evaluation are illustrated under experimental boundary conditions forcing the appearance of oscillations. We demonstrate that the effective pore water volume and the pumping rate can be determined from a conservative tracer breakthrough curve uniquely. In this way, external preconditioning of the material, e.g., drying, can be avoided. A reactive breakthrough experiment revealed a significant increase in the pore water pH value as a consequence of the closed loop. These results highlight the specific impact of the closed mass balance. Furthermore, the basis for the modeling of closed-flow experiments is given by the derivation of constitutive equations and numerical implementation, validated with the presented experiments.

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)

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

Experiments and analysis of thermal stresses around the nozzle of the reactor vessel

International Nuclear Information System (INIS)

Song, D.H.; Oh, J.H.; Song, H.K.; Park, D.S.; Shon, K.H.

1981-01-01

This report describes the results of analysis and experiments on the thermal stress around the reactor vessel nozzle performed to establish a capability of thermal stress analysis of pressure vessel subjected to thermal loadings. Firstly, heat conduction analysis during reactor design transients and analysis on the experimental model were performed using computer code FETEM-1 for the purpose of verification of FETEM-1 which was developed in 1979 and will be used to obtain the temperature distribution in a solid body under the steady-state and the transient conditions. The results of the analysis was compared to the results in the Stress Report of Kori-1 reactor vessel and those from experiments on the model, respectively

Molten material-containing vessel

International Nuclear Information System (INIS)

Akagawa, Katsuhiko

1998-01-01

The molten material-containing vessel of the present invention comprises a vessel main body having an entrance opened at the upper end, a lid for closing the entrance, an outer tube having an upper end disposed at the lower surface of the lid, extended downwardly and having an closed lower end and an inner tube disposed coaxially with the outer tube. When a molten material is charged from the entrance to the inside of the vessel main body of the molten material-containing vessel and the entrance is closed by the lid, the outer tube and the inner tube are buried in the molten material in the vessel main body, accordingly, a fluid having its temperature elevated by absorption of the heat of the molten material rises along the inner circumferential surface of the outer tube, abuts against the lower surface of the lid and cooled by exchanging heat with the lid and forms a circulating flow. Since the heat in the molten material is continuously absorbed by the fluid, transferred to the lid and released from the lid to the atmospheric air, heat releasing efficiency can be improved compared with conventional cases. (N.H.)

FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

Directory of Open Access Journals (Sweden)

Thiollay Nicolas

2016-01-01

Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

Explosion properties of aluminum/oxidizer mixtures in a closed vessel; Aluminium/sankazai kongobutsu no mippei yokinai deno bakuhatsu tokusei

Energy Technology Data Exchange (ETDEWEB)

Miyake, A.; Aochi, T.; Shiraki, K.; Ogawa, T. [Yokohama National University, Yokohama (Japan). Faculty of Engineering

1995-10-31

In order to understand explosion properties of aluminum/oxidized mixtures for firework a closed vessel test was carried out, and pressure profile and ignition delay time were measured. It was found that flake aluminum (Al(f))/oxidized mixtures were more reactive and showed higher pressure values than those of atomized aluminum (Al(a))/oxidized mixtures. At a positive oxygen balance region Al(f)/oxidized mixtures showed a good agreement with the theoretically predicted values of combustion pressure at a constant volume. The combustion parameters with equilibrium calculation of aluminum/potassium chlorate showed almost the same ones of aluminum/potassium perchlorate. Furthermore, to investigate the influence of particle diameter of potassium chlorate on the explosion properties of Al(f)/potassium chlorate mixtures, the same kind of closed vessel test was performed and it was found that the mixtures became less sensitive and reactive with the increase of particle diameter. 17 refs., 7 figs., 2 tabs.

Computational methods for fracture analysis of heavy-section steel technology (HSST) pressure vessel experiments

International Nuclear Information System (INIS)

Bass, B.R.; Bryan, R.H.; Bryson, J.W.; Merkle, J.G.

1983-01-01

This paper summarizes the capabilities and applications of the general-purpose and special-purpose computer programs that have been developed for use in fracture mechanics analyses of HSST pressure vessel experiments. Emphasis is placed on the OCA/USA code, which is designed for analysis of pressurized-thermal-shock (PTS) conditions, and on the ORMGEN/ADINA/ORVIRT system which is used for more general analysis. Fundamental features of these programs are discussed, along with applications to pressure vessel experiments

Review on experiments relating to primary containment vessel failure

International Nuclear Information System (INIS)

Suzuki, Hiroyuki; Okada, Hidetoshi; Uchida, Sunsuke; Naitoh, Masanori

2015-01-01

Experiments regarding failures of primary containment vessels (PCVs) are reviewed and remained issues to be investigated in the future are discussed. Experiments are categorized as those relating to criteria of PCV failures and to FP releases through breaches on PCV boundaries. In the experiments categorized as those relating to criteria of PCV failures, experiments with full-scale, scale models, and compounds used for sealing are surveyed. Experiments relating to an amount of radioactive fission products (FPs) trapped at breaches on PCV boundaries are also reviewed. As remained issues to be investigated in the future, two items are pointed out: Evaluating degradation behavior of PCV boundaries exposed to temperature and pressure from the failure onset criteria to far above them, and evaluating an amount of FPs trapped at breaches on PCV boundaries. (author)

Reactor vessel stud closure system

International Nuclear Information System (INIS)

Spiegelman, S.R.; Salton, R.B.; Beer, R.W.; Malandra, L.J.; Cognevich, M.L.

1982-01-01

A quick-acting stud tensioner apparatus for enabling the loosening or tightening of a stud nut on a reactor vessel stud. The apparatus is adapted to engage the vessel stud by closing a gripper around an upper end of the vessel stud when the apparatus is seated on the stud. Upon lifting the apparatus, the gripper releases the vessel stud so that the apparatus can be removed

Computational methods for fracture analysis of heavy-section steel technology (HSST) pressure vessel experiments

International Nuclear Information System (INIS)

Bass, B.R.; Bryan, R.H.; Bryson, J.W.; Merkle, J.G.

1985-01-01

This paper summarizes the capabilities and applications of the general-purpose and special-purpose computer programs that have been developed at ORNL for use in fracture mechanics analyses of HSST pressure vessel experiments. Emphasis is placed on the OCA/USA code, which is designed for analysis of pressurized-thermal-shock (PTS) conditions, and on the ORMGEN/ADINA/ORVIRT system which is used for more general analysis. Fundamental features of these programs are discussed, along wih applications to pressure vessel experiments. (orig./HP)

Pressurized wet digestion in open vessels (T11)

International Nuclear Information System (INIS)

Kettisch, P.; Maichin, P.; Zischka, M.; Knapp, G.

2002-01-01

Full text: Pressurized wet digestion in closed vessels, microwave assisted or with conventional conductive heating, is the most important sample preparation technique for digestion or leaching procedures in element analysis. In comparison to open vessel digestion closed vessel digestion methods have many advantages, but there is one disadvantage - complex and expensive vessel designs. A new technique - pressurized wet digestion in open vessels - combine the advantages of closed vessel sample digestion with the application of simple and cheap open vessels made of quartz or PFA. The vessels are placed in a high pressure Asher HPA, which is adapted with a Teflon liner and filled partly with water. The analytical results with 30 ml quartz vessels, 22 ml PFA vessels and 1.5 ml PIA auto sampler cups will be shown. In principle every dimensions of vessels can be used. The vessels are loaded with sample material (max. 1.5 g with quartz vessels, max. 0.5 g with PFA vessels and 50 mg with auto sampler cups) and digestion reagent. Afterwards the vessels are simply covered with PTFE stoppers and not sealed. The vessels are transferred into a special adapted HPA and digested at temperatures up to 270 o C. The digestion time is 90 min. and cooling down to room temperature 30 min. The analytical results of CRM's are within the certified values and no cross contamination and losses of volatile elements could be observed. (author)

Mobile nuclear reactor containment vessel

International Nuclear Information System (INIS)

Thompson, R.E.; Spurrier, F.R.; Jones, A.R.

1978-01-01

A containment vessel for use in mobile nuclear reactor installations is described. The containment vessel completely surrounds the entire primary system, and is located as close to the reactor primary system components as is possible in order to minimize weight. In addition to being designed to withstand a specified internal pressure, the containment vessel is also designed to maintain integrity as a containment vessel in case of a possible collision accident

Automatic Vessel Segmentation on Retinal Images

Institute of Scientific and Technical Information of China (English)

Chun-Yuan Yu; Chia-Jen Chang; Yen-Ju Yao; Shyr-Shen Yu

2014-01-01

Several features of retinal vessels can be used to monitor the progression of diseases. Changes in vascular structures, for example, vessel caliber, branching angle, and tortuosity, are portents of many diseases such as diabetic retinopathy and arterial hyper-tension. This paper proposes an automatic retinal vessel segmentation method based on morphological closing and multi-scale line detection. First, an illumination correction is performed on the green band retinal image. Next, the morphological closing and subtraction processing are applied to obtain the crude retinal vessel image. Then, the multi-scale line detection is used to fine the vessel image. Finally, the binary vasculature is extracted by the Otsu algorithm. In this paper, for improving the drawbacks of multi-scale line detection, only the line detectors at 4 scales are used. The experimental results show that the accuracy is 0.939 for DRIVE (digital retinal images for vessel extraction) retinal database, which is much better than other methods.

Sample preparation techniques based on combustion reactions in closed vessels - A brief overview and recent applications

International Nuclear Information System (INIS)

Flores, Erico M.M.; Barin, Juliano S.; Mesko, Marcia F.; Knapp, Guenter

2007-01-01

In this review, a general discussion of sample preparation techniques based on combustion reactions in closed vessels is presented. Applications for several kinds of samples are described, taking into account the literature data reported in the last 25 years. The operational conditions as well as the main characteristics and drawbacks are discussed for bomb combustion, oxygen flask and microwave-induced combustion (MIC) techniques. Recent applications of MIC techniques are discussed with special concern for samples not well digested by conventional microwave-assisted wet digestion as, for example, coal and also for subsequent determination of halogens

Fine-grained visual marine vessel classification for coastal surveillance and defense applications

Science.gov (United States)

Solmaz, Berkan; Gundogdu, Erhan; Karaman, Kaan; Yücesoy, Veysel; Koç, Aykut

2017-10-01

The need for capabilities of automated visual content analysis has substantially increased due to presence of large number of images captured by surveillance cameras. With a focus on development of practical methods for extracting effective visual data representations, deep neural network based representations have received great attention due to their success in visual categorization of generic images. For fine-grained image categorization, a closely related yet a more challenging research problem compared to generic image categorization due to high visual similarities within subgroups, diverse applications were developed such as classifying images of vehicles, birds, food and plants. Here, we propose the use of deep neural network based representations for categorizing and identifying marine vessels for defense and security applications. First, we gather a large number of marine vessel images via online sources grouping them into four coarse categories; naval, civil, commercial and service vessels. Next, we subgroup naval vessels into fine categories such as corvettes, frigates and submarines. For distinguishing images, we extract state-of-the-art deep visual representations and train support-vector-machines. Furthermore, we fine tune deep representations for marine vessel images. Experiments address two scenarios, classification and verification of naval marine vessels. Classification experiment aims coarse categorization, as well as learning models of fine categories. Verification experiment embroils identification of specific naval vessels by revealing if a pair of images belongs to identical marine vessels by the help of learnt deep representations. Obtaining promising performance, we believe these presented capabilities would be essential components of future coastal and on-board surveillance systems.

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

Molten core debris-sodium interactions: M-Series experiments

International Nuclear Information System (INIS)

Sowa, E.S.; Gabor, J.D.; Pavlik, J.R.; Cassulo, J.C.; Cook, C.J.; Baker, L. Jr.

1979-01-01

Five new kilogram-scale experiments have been carried out. Four of the experiments simulated the situation where molten core debris flows from a breached reactor vessel into a dry reactor cavity and is followed by a flow of sodium (Ex-vessel case) and one experiment simulated the flow of core debris into an existing pool of sodium (In-vessel case). The core debris was closely simulated by a thermite reaction which produced a molten mixture of UO 2 , ZrO 2 , and stainless steel. There was efficient fragmentation of the debris in all experiments with no explosive interactions observed

Corroboration of dynamic characteristics of FBR main vessels by pseudo-dynamic and dynamic buckling experiments

International Nuclear Information System (INIS)

Kokubo, K.; Nakagawa, M.; Kawamoto, Y.; Murakami, T.; Matuura, S.; Hagiwara, Y.

1991-01-01

Shaking table tests for small-scale models and pseudo-dynamic buckling tests for moderate-scale models are conducted in order to investigate nonlinear pre- and post-buckling characteristics of fast breeder reactor vessels under the seismic lateral load. Two types of ground acceleration waves are used in the experiments. Nonlinear one-degree-of-freedom numerical simulations are also conducted using the hysteresis rules obtained by the tests. Good agreements are obtained between the experiments and calculations. The design method for vessels based on the estimation of nonlinear buckling behaviors is considered. (author)

Depressurization as a means of leak checking large vacuum vessels

International Nuclear Information System (INIS)

Callis, R.W.; Langhorn, A.; Petersen, P.I.; Ward, C.; Wesley, J.

1985-01-01

A common problem associated with large vacuum vessels used in magnetic confinement fusion experiments is that leak checking is hampered by the inaccessibility to most of the vacuum vessel surface. This inaccessibility is caused by the close proximity of magnetic coils, diagnostics and, for those vessels that are baked, the need to completely surround the vessel with a thermal insulation blanket. These obstructions reduce the effectiveness of the standard leak checking method of using a mass spectrometer and spraying a search gas such as helium on the vessel exterior. Even when the presence of helium is detected, its entry point into the vessel cannot always be pinpointed. This paper will describe a method of overcoming this problem. By slightly depressurizing the vessel, an influx of helium through the leak is created. The leak site can then be identified by personnel within the vessel using standard sniffing procedures. There are two conditions which make this method of leak checking practical. First, the vessel need only be depressurized 2 psi, thus allowing personnel inside to perform the sniffing operation. Second, the sniffing probe used (Leybold--Heraus ''Quick Test'') could detect a change in helium concentration as small as 100 ppb, which allows for faster scanning of the vessel inferior. Use of this technique to find an elusive 10 -3 Torrxl/s leak in the Doublet III tokamak vacuum vessel will be presented

Radioactive waste processing vessel

International Nuclear Information System (INIS)

Hayashi, Masaru; Suzuki, Osamu; Ishizaki, Kanjiro.

1987-01-01

Purpose: To obtain a vessel of a reduced weight and with no external leaching of radioactive materials. Constitution: The vessel main body is constituted, for example, with light weight concretes or foamed concretes, particularly, foamed concretes containing fine closed bubbles in the inside. Then, layers having dense texture made of synthetic resin such as polystylene, vinylchloride resin, etc. or metal plate such as stainless plate are integrally disposed to the inner surface of the vessel main body. The cover member also has the same structure. (Sekiya, K.)

Sodium oxide aerosol behavior in a closed vessel. Comparison of computer modeling with aerosol experiments

International Nuclear Information System (INIS)

Fermandjian, Jean.

1979-08-01

Fast breeder reactor safety needs models validated to predict the behavior of sodium aerosols in the different reactor compartments during hypothetical sodium accident. Besides their chemical toxicity, the sodium aerosols are a transfer vector of radioactivity during a contaminated sodium fire. The purpose of this work is to validate models (HAARM 2 and PARDISEKO 3) with tests of sodium pool fires in a 400 m 3 concrete vessel in a confined atmosphere (CASSANDRE tests). The comparison between calculations and experimental results reveals that difficulties still exist, especially as to the selection of the values to be given to some input parameters (physical data of experimental origin, in particular the aerosols source function, the characteristics of the distribution of the emitted particles and the form factor of the agglomerated particles) [fr

Design, fabrication and operating experience of Monju ex-vessel fuel storage tank

International Nuclear Information System (INIS)

Yokota, Yoshio; Yamagishi, Yoshiaki; Kuroha, Mitsuo; Inoue, Tatsuya

1995-01-01

In FBRs there are two methods of storing and cooling the spent fuel - the in-vessel storage and the ex-vessel storage. Because of the sodium leaks through the tank at the beginning of pre-operation, the utilization of the ex-vessel fuel storage tank (EVST) of some FBR plant has been changed from the ex-vessel fuel storage to the interim fuel transfer tank. This led to reactor designers focusing on the material, structure and fabrication of the carbon steel sodium storage tanks worldwide. The Monju EVST was at the final stage of the design, when the leaks occurred. The lesson learned from that experience and the domestic fabrication technology are reflected to the design and fabrication of the Monju EVST. This paper describes the design, fabrication and R and D results for the tank, and operating experience in functional test. The items to be examined are as follows: (1) Overall structure of the tank and design philosophy on the function, (2) Structure of the cover shielding plug and its design philosophy, (3) Structures of the rotating rack and its bearings, and their design philosophy, (4) Cooling method and its design philosophy, (5) Structure and fabrication of the cooling coil support inside EVST with comparison of leaked case, (6) R and D effort for items above. The fabrication of the Monju EVST started in August 1986 and it was shipped to the site in March 1990. Installation was completed in November 1990, and sodium fill after pre-heating started in 1991. The operation has been continued since September 1992. In 1996 when the first spent fuel is stored, its total functions will be examined. (author)

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Concrete benchmark experiment: ex-vessel LWR surveillance dosimetry; Experience ``Benchmark beton`` pour la dosimetrie hors cuve dans les reacteurs a eau legere

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, H.; D`Hondt, P.; Oeyen, J.; Risch, P.; Bioux, P.

1993-09-01

The analysis of DOEL-1 in-vessel and ex-vessel neutron dosimetry, using the DOT 3.5 Sn code coupled with the VITAMIN-C cross-section library, showed the same C/E values for different detectors at the surveillance capsule and the ex-vessel cavity positions. These results seem to be in contradiction with those obtained in several Benchmark experiments (PCA, PSF, VENUS...) when using the same computational tools. Indeed a strong decreasing radial trend of the C/E was observed, partly explained by the overestimation of the iron inelastic scattering. The flat trend seen in DOEL-1 could be explained by compensating errors in the calculation such as the backscattering due to the concrete walls outside the cavity. The `Concrete Benchmark` experiment has been designed to judge the ability of this calculation methods to treat the backscattering. This paper describes the `Concrete Benchmark` experiment, the measured and computed neutron dosimetry results and their comparison. This preliminary analysis seems to indicate an overestimation of the backscattering effect in the calculations. (authors). 5 figs., 1 tab., 7 refs.

Helium leak testing of large pressure vessels or subassemblies

International Nuclear Information System (INIS)

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

1977-01-01

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

Pressure vessel design manual

CERN Document Server

Moss, Dennis R

2013-01-01

Pressure vessels are closed containers designed to hold gases or liquids at a pressure substantially different from the ambient pressure. They have a variety of applications in industry, including in oil refineries, nuclear reactors, vehicle airbrake reservoirs, and more. The pressure differential with such vessels is dangerous, and due to the risk of accident and fatality around their use, the design, manufacture, operation and inspection of pressure vessels is regulated by engineering authorities and guided by legal codes and standards. Pressure Vessel Design Manual is a solutions-focused guide to the many problems and technical challenges involved in the design of pressure vessels to match stringent standards and codes. It brings together otherwise scattered information and explanations into one easy-to-use resource to minimize research and take readers from problem to solution in the most direct manner possible. * Covers almost all problems that a working pressure vessel designer can expect to face, with ...

Irradiation embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

Bros, J.

2000-01-01

From the historical decision of closing the Yankee Rowe NPP because of the uncertainties on the level of reactor pressure vessel neutron embrittlement, this paper reviews the technical-scientist bases of the degradation phenomena, and refers to the evolution of reactor pressure vessel radiation surveillance programs. (Author)

High pressure deuterium-tritium gas target vessels for muon-catalyzed fusion experiments

International Nuclear Information System (INIS)

Caffrey, A.J.; Spaletta, H.W.; Ware, A.G.; Zabriskie, J.M.; Hardwick, D.A.; Maltrud, H.R.; Paciotti, M.A.

1989-01-01

In experimental studies of muon-catalyzed fusion, the density of the hydrogen gas mixture is an important parameter. Catalysis of up to 150 fusions per muon has been observed in deuterium-tritium gas mixtures at liquid hydrogen density; at room temperature, such densities require a target gas pressure of the order of 1000 atmospheres (100 MPa, 15,000 psi). We report here the design considerations for hydrogen gas target vessels for muon-catalyzed fusion experiments that operate at 1000 and 10,000 atmospheres. The 1000 atmosphere high pressure target vessels are fabricated of Type A-286 stainless steel and lined with oxygen-free, high-conductivity (OFHC) copper to provide a barrier to hydrogen permeation of the stainless steel. The 10,000 atmosphere ultrahigh pressure target vessels are made from 18Ni (200 grade) maraging steel and are lined with OFHC copper, again to prevent hydrogen permeation of the steel. In addition to target design features, operating requirements, fabrication procedures, and secondary containment are discussed. 13 refs., 3 figs., 1 tab

Natural Convection Heat Transfer of Oxide Pool During In-Vessel Retention of Core Melts

Energy Technology Data Exchange (ETDEWEB)

Park, Hae-Kyun; Chung, Bum-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The integrity of reactor vessel may be threatened by the heat generation at the oxide pool and to the natural convection heat transfer to the reactor vessel by those two layers. Therefore, External Reactor Vessel Cooling (ERVC) is performed in order to secure the integrity of the reactor vessel. Whether the IVR(In-Vessel Retention) Strategy can be applicable to a larger reactor is the technical concern, which nourished the research interest for the natural convection heat transfer of metal and oxide pool and ERVC performance. Especially, it is hard to simulate oxide pool by experimentally due to the high level of buoyancy. Moreover, the volumetrically exothermic working fluid should be adopted to simulate the behavior of the core melts. Therefore, the volumetric heat sources that immersed in the working fluid have been adopted to simulate oxide pool by experiment. We investigated oxide pool with two different designs of the volumetric heat sources that adopted previous experiments. The investigation was performed by mass transfer experiment using analogy between heat and mass transfers. The results were compared to previous studies. We simulated the natural convection heat transfer of the oxide pool by mass transfer experiment. The isothermally cooled condition was established by limiting current technique firstly. The results were compared to previous studies under identical design of the volumetric heat sources. The average Nu's of the curvature and the top plate were close to the previous studies.

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

International Nuclear Information System (INIS)

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

1976-08-01

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

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

Nuclear reactor vessel decontamination systems

International Nuclear Information System (INIS)

McGuire, P. J.

1985-01-01

There is disclosed in the present application, a decontamination system for reactor vessels. The system is operatable without entry by personnel into the contaminated vessel before the decontamination operation is carried out and comprises an assembly which is introduced into the vertical cylindrical vessel of the typical boiling water reactor through the open top. The assembly includes a circular track which is centered by guideways permanently installed in the reactor vessel and the track guides opposed pairs of nozzles through which water under very high pressure is directed at the wall for progressively cutting and sweeping a tenacious radioactive coating as the nozzles are driven around the track in close proximity to the vessel wall. The whole assembly is hoisted to a level above the top of the vessel by a crane, outboard slides on the assembly brought into engagement with the permanent guideways and the assembly progressively lowered in the vessel as the decontamination operation progresses. The assembly also includes a low pressure nozzle which forms a spray umbrella above the high pressure nozzles to contain radioactive particles dislodged during the decontamination

Cardiac and great vessel injuries after chest trauma: our 10-year experience.

Science.gov (United States)

Onan, Burak; Demirhan, Recep; Öz, Kürşad; Onan, Ismihan Selen

2011-09-01

Cardiovascular injuries after trauma present with high mortality. The aim of the study was to present our experience in cardiac and great vessel injuries after chest trauma. During the 10-year period, 104 patients with cardiac (n=94) and great vessel (n=10) injuries presented to our hospital. The demographic data, mechanism of injury, location of injury, other associated injuries, timing of surgical intervention, surgical approach, and clinical outcome were reviewed. Eighty-eight (84.6%) males presented after chest trauma. The mean age of the patients was 32.5±8.2 years (range: 12-76). Penetrating injuries (62.5%) were the most common cause of trauma. Computed tomography was performed in most cases and echocardiography was used in some stable cases. Cardiac injuries mostly included the right ventricle (58.5%). Great vessel injuries involved the subclavian vein in 6, innominate vein in 1, vena cava in 1, and descending aorta in 2 patients. Early operations after admission to the emergency were performed in 75.9% of the patients. Thoracotomy was performed in 89.5% of the patients. Operative mortality was significantly high in penetrating injuries (p=0.01). Clinicians should suspect cardiac and great vessel trauma in every patient presenting to the emergency unit after chest trauma. Computed tomography and echocardiography are beneficial in the management of chest trauma. Operative timing depends on hemodynamic status, and a multidisciplinary team approach improves the patient's prognosis.

Method for the radiographic examination of the walls or components of an essentially closed vessel, and also the provision of means for the application of the method

International Nuclear Information System (INIS)

1978-01-01

Method for the radiographic examination of the wall ports or supporting components of an essentially closed vessel, whereby one brings to the side of the vessel walls or supports under examination a radiation source and, to the opposite side, a radiation sensitive film, the film being irradiated by the source and thereafter developed, characterised in that one introduces into the inside of the vessel a hollow tube at a unique distance from the wall or support component, at least one end of the hollow tube being fed out and in which the hollow tube, during the period of the examination, the irradiation source or an irradiation sensitive film is introduced. (G.C.)

Study on closed pressure vessel test. Effect of heat rate, sample weight and vessel size on pressure rise due to thermal decomposition; Mippeigata atsuryoku yoki shiken ni kansuru kenkyu. Atsuryoku hassei kyodo ni oyobosu kanetsusokudo, shiryoryo oyobi youki saizu no eikyo

Energy Technology Data Exchange (ETDEWEB)

Aoki, Kenji.; Akutsu, Yoshiaki.; Arai, Mitsuru.; Tamura, Masamitsu. [The University of Tokyo, Tokyo (Japan). School of Engineering

1999-02-28

We have attempted to devise a new closed pressure vessel test apparatus in order to evaluate the violence of thermal decomposition of self-reactive materials and have examined some influencing factors, such as heat rate, sample weight, filling factor (sample weight/vessel size) and vessel size on Pmax (maximum pressure rise) and dP/dt (rate of pressure rise) due to their thermal decomposition. As a result, the following decreasing orders of Pmax and dP/dt were shown. Pmax: ADCA>BPZ>AIBN>TCP dP/dt: AIBN>BPZ>ADCA>TCP Moreover, Pmax was not almost influenced by heat rate, while dP/dt increased with an increase in heat rate in the case of BPZ. Pmax and dP/dt increased with an increase in sample weight and the degree of increase depended on the kinds of materials. In addition, it was shown that Pmax and dP/dt increased with an increase in vessel size at a constant filling factor. (author)

Safety Research Experiment Facility Project. Conceptual design report. Volume V. Reactor vessel and closure

International Nuclear Information System (INIS)

1975-12-01

The Prestressed Concrete Reactor Vessel (PCRV) will serve as the primary pressure retaining structure for the Safety Research Experiment Facility (SAREF) reactor. The reactor core, control rod drive room, primary heat exchangers, and gas circulators will be located in cavities within the PCRV. The orientation of these cavities, except for the control rod drive room, will be similar to the high-temperature gas-cooled reactor (HTGR) designs that are currently proposed or under design. Due to the nature of this type of structure, all biological and radiological shielding requirements are incorporated into the basic vessel design. At the midcore plane there are three radially oriented slots that will extend from the outside surface of the PCRV to the reactor core liner. These slots will accommodate each of the fuel motion monitoring systems which will be part of the observation apparatus used with the loop experiments

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)

Experiments for habitation-related technologies in closed system

International Nuclear Information System (INIS)

Masuda, Tsuyoshi; Tako, Yasuhiro

2006-01-01

Preliminary habitation experiments, including facility implementation tests, investigation of work load and psychological and physiological effect of human subjects, environmental monitoring tests, and integrated implementation tests were performed in Closed Ecology Experiment Facilities (CEEF). Results showed insufficient ability of CO 2 separator, issues of overtime work of subjects and inadequate communication between subjects and supporting staff outside the modules. Countermeasures for these issues were developed. Results show that habitation experiments from fiscal year 2005 on the CEEF were judged to be feasible. (author)

Ionizing radiations and blood vessels

International Nuclear Information System (INIS)

Vorob'ev, E.I.; Stepanov, R.P.

1985-01-01

Data on phenomeology of radiation changes of blood vessels are systemized and the authors' experience is generalyzed. A critical analysis of modern conceptions on processes resulting in vessel structure damage after irradiation, is given. Special attention is paid to reparation and compensation of radiation injury of vessels

The TPX vacuum vessel and in-vessel components

International Nuclear Information System (INIS)

Heitzenroeder, P.; Bialek, J.; Ellis, R.; Kessel, C.; Liew, S.

1994-01-01

The Tokamak Physics Experiment (TPX) is a superconducting tokamak with double-null diverters. TPX is designed for 1,000-second discharges with the capability of being upgraded to steady state operation. High neutron yields resulting from the long duration discharges require that special consideration be given to materials and maintainability. A unique feature of the TPX is the use of a low activation, titanium alloy vacuum vessel. Double-wall vessel construction is used since it offers an efficient solution for shielding, bakeout and cooling. Contained within the vacuum vessel are the passive coil system, Plasma Facing Components (PFCs), magnetic diagnostics, and the internal control coils. All PFCs utilize carbon-carbon composites for exposed surfaces

Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

International Nuclear Information System (INIS)

Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

1984-01-01

Results of reactor-material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address ex-vessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debris characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity

Irradiation experiments on materials for core internals, pressure vessel and fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Tsukada, Takashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

Materials degradation due to the aging phenomena is one of the key issues for the life assessment and extension of the light water reactors (LWRs). This presentation introduces JAERI`s activities in the field of LWR material researches which utilize the research and testing reactors for irradiation experiments. The activities are including the material studies for the core internals, pressure vessel and fuel cladding. These materials are exposed to the neutron/gamma radiation and high temperature water environments so that it is worth reviewing their degradation phenomena as the continuum. Three topics are presented; For the core internal materials, the irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steels is the present major concern. At JAERI the effects of alloying elements on IASCC have been investigated through the post-irradiation stress corrosion cracking tests in high-temperature water. The radiation embrittlement of pressure vessel steels is still a significant issue for LWR safety, and at JAERI some factors affecting the embrittlement behavior such as a dose rate have been investigated. Waterside corrosion of Zircaloy fuel cladding is one of the limiting factors in fuel rod performance and an in-situ measurement of the corrosion rate in high-temperature water was performed in JMTR. To improve the reliability of experiments and to extent the applicability of experimental techniques, a mutual utilization of the technical achievements in those irradiation experiments is desired. (author)

Ionizing radiations and blood vessels

International Nuclear Information System (INIS)

Vorob'ev, E.I.; Stepanov, R.P.

1985-01-01

Data on phenomenology of radiation-induced changes in blood vessels are systematized and authors' experience is generalized. Modern concepts about processes leading to vessel structure injury after irradiation is critically analyzed. Special attention is paid to reparation and compensation of X-ray vessel injury, consideration of which is not yet sufficiently elucidated in literature

Method of transferring regular shaped vessel into cell

International Nuclear Information System (INIS)

Murai, Tsunehiko.

1997-01-01

The present invention concerns a method of transferring regular shaped vessels from a non-contaminated area to a contaminated cell. A passage hole for allowing the regular shaped vessels to pass in the longitudinal direction is formed to a partitioning wall at the bottom of the contaminated cell. A plurality of regular shaped vessel are stacked in multiple stages in a vertical direction from the non-contaminated area present below the passage hole, allowed to pass while being urged and transferred successively into the contaminated cell. As a result, since they are transferred while substantially closing the passage hole by the regular shaped vessels, radiation rays or contaminated materials are prevented from discharging from the contaminated cell to the non-contaminated area. Since there is no requirement to open/close an isolation door frequently, the workability upon transfer can be improved remarkably. In addition, the sealing member for sealing the gap between the regular shaped vessel passing through the passage hole and the partitioning wall of the bottom is disposed to the passage hole, the contaminated materials in the contaminated cells can be prevented from discharging from the gap to the non-contaminated area. (N.H.)

Closed Loop Experiment Manager (CLEM—An Open and Inexpensive Solution for Multichannel Electrophysiological Recordings and Closed Loop Experiments

Directory of Open Access Journals (Sweden)

Hananel Hazan

2017-10-01

Full Text Available There is growing need for multichannel electrophysiological systems that record from and interact with neuronal systems in near real-time. Such systems are needed, for example, for closed loop, multichannel electrophysiological/optogenetic experimentation in vivo and in a variety of other neuronal preparations, or for developing and testing neuro-prosthetic devices, to name a few. Furthermore, there is a need for such systems to be inexpensive, reliable, user friendly, easy to set-up, open and expandable, and possess long life cycles in face of rapidly changing computing environments. Finally, they should provide powerful, yet reasonably easy to implement facilities for developing closed-loop protocols for interacting with neuronal systems. Here, we survey commercial and open source systems that address these needs to varying degrees. We then present our own solution, which we refer to as Closed Loop Experiments Manager (CLEM. CLEM is an open source, soft real-time, Microsoft Windows desktop application that is based on a single generic personal computer (PC and an inexpensive, general-purpose data acquisition board. CLEM provides a fully functional, user-friendly graphical interface, possesses facilities for recording, presenting and logging electrophysiological data from up to 64 analog channels, and facilities for controlling external devices, such as stimulators, through digital and analog interfaces. Importantly, it includes facilities for running closed-loop protocols written in any programming language that can generate dynamic link libraries (DLLs. We describe the application, its architecture and facilities. We then demonstrate, using networks of cortical neurons growing on multielectrode arrays (MEA that despite its reliance on generic hardware, its performance is appropriate for flexible, closed-loop experimentation at the neuronal network level.

Rapid construction of concrete pressure vessels

International Nuclear Information System (INIS)

Limbert, D.; Weatherseed, D.C.

1989-01-01

This paper opens with a general description of the concrete pressure vessel followed by a more detailed examination of the critical elements of the construction, including choice of methods and plant which were selected to ensure its rapid construction. The pressure vessel construction cannot be treated in isolation, because it is very closely linked with its surrounding structures - namely the reactor hall which surrounds it and the charge hall which tops it, as will be seen in the context of this paper. Rate of progress of construction is not entirely in the civil contractor's hands because so many of the operations affecting the civil works are of a mechanical nature, hence a very close liaison and understanding amongst all contractors concerned was of the utmost importance. (author)

Experiments for neutron fluence assessment on WWER-440 and WWER-1000 pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Ilieva, K; Apostolov, T; Penev, I; Trifonov, A; Taskaev, E; Belousov, S; Antonov, S; Petrova, T; Stoeva, L [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika; Boyadzhiev, Z; Nelov, N; Tsocheva, V; Andreeva, I; Lilkov, B; Velichkov, V; Monev, M [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

1996-12-31

The activity of shavings sampled out from the expected maximum embrittlement location (weld 4) on the inner pressure vessel wall of the Kozloduy-1 Unit after the 14-th cycle has been measured. The experiment was carried out along the INEI channel using Fe and Cu string and foil detectors. The axial neutron flux distribution at the Unit 3 after the cycle 11 has been measured and compared to the calculated values. The calculations of the expected activities have been carried out taking into account the local power distribution. A comparison between measured and calculated values using ACTIVAT code is made. It shows a discrepancy of about 20%. It is recommended to carry out ex-vessel neutron fluence measurements using a rack device with activation detectors in order to verify the calculation results. 8 refs., 3 figs., 2 tabs.

Transformations to diagonal bases in closed-loop quantum learning control experiments

International Nuclear Information System (INIS)

Cardoza, David; Trallero-Herrero, Carlos; Langhojer, Florian; Rabitz, Herschel; Weinacht, Thomas

2005-01-01

This paper discusses transformations between bases used in closed-loop learning control experiments. The goal is to transform to a basis in which the number of control parameters is minimized and in which the parameters act independently. We demonstrate a simple procedure for testing whether a unitary linear transformation (i.e., a rotation amongst the control variables) is sufficient to reduce the search problem to a set of globally independent variables. This concept is demonstrated with closed-loop molecular fragmentation experiments utilizing shaped, ultrafast laser pulses

An international survey of in-service inspection experience with prestressed concrete pressure vessels and containments for nuclear reactors

International Nuclear Information System (INIS)

1982-04-01

An international survey is presented of experience obtained from the in-service surveillance of prestressed concrete pressure vessels and containments for nuclear reactors. Some information on other prestressed concrete structures is also given. Experience has been gained during the working life of such structures in Western Europe and the USA over the years since 1967. For each country a summary is given of the nuclear programme, national standards and Codes of Practice, and the detailed in-service inspection programme. Reports are then given of the actual experience obtained from the inspection programme and the methods of measurement, examination and reporting employed in each country. A comprehensive bibliography of over 100 references is included. The appendices contain information on nuclear power stations which are operating, under construction or planned worldwide and which employ either prestressed concrete pressure vessels or containments. (U.K.)

Fracture-mechanics data deduced from thermal-shock and related experiments with LWR pressure-vessel material

International Nuclear Information System (INIS)

Cheverton, R.D.; Canonico, D.A.; Iskander, S.K.; Bolt, S.E.; Holz, P.P.; Nanstad, R.K.; Stelzman, W.J.

1982-01-01

Pressurized water reactors (PWRs) are susceptible to certain types of hypothetical accidents that can subject the reactor pressure vessel to severe thermal shock, that is, a rapid cooling of the inner surface of the vessel wall. The thermal-shock loading, coupled with the radiation-induced reduction in the material fracture toughness, introduces the possibility of propagation of preexistent flaws and what at one time were regarded as somewhat unique fracture-oriented conditions. Several postulated reactor accidents have been analyzed to discover flaw behavior trends; seven intermediate-scale thermal-shock experiments with steel cylinders have been conducted; and corresponding materials characterization studies have been performed. Flaw behavior trends and related fracture-mechanics data deduced from these studies are discussed

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.

Brain death: close relatives' use of imagery as a descriptor of experience.

Science.gov (United States)

Frid, Ingvar; Haljamäe, Hengo; Ohlén, Joakim; Bergbom, Ingegerd

2007-04-01

This paper is a report of a study to explore the use of imagery to describe the experience of confronting brain death in a close relative. The brain death of a loved one has been described as an extremely difficult experience for close relatives, evoking feelings of anger, emotional pain, disbelief, guilt and suffering. It can also be difficult for relatives to distinguish brain death from the state of coma and thus difficult to apprehend information about the diagnosis. Narrative theory and a hermeneutic phenomenological method guided the interpretation of 17 narratives from close relatives of brain dead patients. All narratives were scrutinized for experiences of brain death. Data were primarily collected in 1999. The primary analysis related to close relatives' experience of brain death in a loved one. A secondary analysis of the imagery they used to describe their experience was carried out in 2003. Six categories of imagery used to describe the experience of confronting a diagnosis of brain death in a loved one emerged: chaotic unreality; inner collapse; sense of forlornness; clinging to the hope of survival; reconciliation with the reality of death; receiving care which gives comfort. Participants also identified two pairs of dimensions to describe their feelings about the relationship between their brain dead relative's body and personhood: presence-absence and divisibility-indivisibility. Being confronted with brain death meant entering into the anteroom of death, facing a loved one who is 'living-dead', and experiencing a chaotic drama of suffering. It is very important for members of the intensive care unit team to recognize, face and respond to these relatives' chaotic experiences, which cause them to need affirmation, comfort and caring. Relatives' use of imagery could be the starting point for a caring conversation about their experiences, either in conversations at the time of the death or when relatives are contacted in a later follow-up.

Analysis and optimization on in-vessel inspection robotic system for EAST

International Nuclear Information System (INIS)

Zhang, Weijun; Zhou, Zeyu; Yuan, Jianjun; Du, Liang; Mao, Ziming

2015-01-01

Since China has successfully built her first Experimental Advanced Superconducting TOKAMAK (EAST) several years ago, great interest and demand have been increasing in robotic in-vessel inspection/operation systems, by which an observation of in-vessel physical phenomenon, collection of visual information, 3D mapping and localization, even maintenance are to be possible. However, it has been raising many challenges to implement a practical and robust robotic system, due to a lot of complex constraints and expectations, e.g., high remanent working temperature (100 °C) and vacuum (10"−"3 pa) environment even in the rest interval between plasma discharge experiments, close-up and precise inspection, operation efficiency, besides a general kinematic requirement of D shape irregular vessel. In this paper we propose an upgraded robotic system with redundant degrees of freedom (DOF) manipulator combined with a binocular vision system at the tip and a virtual reality system. A comprehensive comparison and discussion are given on the necessity and main function of the binocular vision system, path planning for inspection, fast localization, inspection efficiency and success rate in time, optimization of kinematic configuration, and the possibility of underactuated mechanism. A detailed design, implementation, and experiments of the binocular vision system together with the recent development progress of the whole robotic system are reported in the later part of the paper, while, future work and expectation are described in the end.

Thermal-buckling analysis of an LMFBR overflow vessel

International Nuclear Information System (INIS)

Severud, L.K.

1983-01-01

During a reactor scram, cold sodium flows into the hot overflow vessel. The effect on the vessel is a compressive thermal stress in a zone just above the sodium level. This condition must be sufficiently controlled to preclude thermal buckling. Also, under repeated scrams, the vessel should not suffer thermal stress low cycle fatigue. To evaluate the closeness to buckling and satisfaction of ASMA Code limits, a combination of simple approximations, detailed elastic shell buckling analyses, and correlations to results of thermal buckling tests were employed. This paper describes the analysis methods, special considerations, and evaluations accomplished for this FFTF vessel to assure satisfaction of ASME buckling design criteria, rules, and limits

Micro Vascular Plug (MVP)-assisted vessel occlusion in neurovascular pathologies: technical results and initial clinical experience.

Science.gov (United States)

Beaty, Narlin B; Jindal, Gaurav; Gandhi, Dheeraj

2015-10-01

Deconstructive approaches may be necessary to treat a variety of neurovascular pathologies. Recently, a new device has become available for endovascular arterial occlusion that may have unique applications in neurovascular disease. The Micro Vascular Plug (MVP, Reverse Medical, Irvine, California, USA) has been designed for vessel occlusion through targeted embolization. To report the results from our initial experience with eight consecutive patients in whom the MVP was used to achieve endovascular occlusion of an artery in the head and neck. Eight consecutive patients treated over a nine-month period were included. The patients' radiographic and electronic medical records were retrospectively reviewed. Specifically demographic information, clinical indication, site of arterial occlusion, size of MVP, time to vessel occlusion, clinical complications, use of other secondary embolic agents, and clinical outcome were recorded. Follow-up information when available is presented. The MVP was used in eight patients for the treatment of neurovascular disease. Indications for treatment included post-traumatic head/neck bleeding (n=3), carotid-cavernous fistula (1), vertebral-vertebral fistula (1), giant fusiform vertebral aneurysm (1), stump-emboli after carotid dissection (1), and iatrogenic vertebral artery penetrating injury (1). One device was used in five patients, two in two patients, and one patient with extensive vertebral-vertebral venous fistula required three plugs to effectively trap the fistula from proximal and distal aspects. Vessel occlusion was obtained in MVP in neurovascular disease. Use of this device may be associated with shorter procedural times and cost savings in comparison with the use of microcoils for vessel occlusion. Our experience shows that MVP can have unique applications in neurovascular pathologies and it complements other occlusive devices. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted

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

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.

Experimental study of in-and-ex-vessel melt cooling during a severe accident

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Baik; Yoo, K J; Park, C K; Seok, S D; Park, R J; Yi, S J; Kang, K H; Ham, Y S; Cho, Y R; Kim, J H; Jeong, J H; Shin, K Y; Cho, J S; Kim, D H

1997-07-01

After code damage during a severe accident in a nuclear reactor, the degraded core has to be cooled down and the decay heat should be removed in order to cease the accident progression and maintain a stable state. The cooling of core melt is divided into in-vessel and ex-vessel cooling depending on the location of molten core which is dependent on the timing of vessel failure. Since the cooling mechanism varies with the conditions of molten core and surroundings and related phenomena, it contains many phenomenological uncertainties so far. In this study, an experimental study for verification of in-vessel corium cooling and several separate effect experiments for ex-vessel cooling are carried out to verify in- and ex-vessel cooling phenomena and finally to develop the accident management strategy and improve engineered reactor design for the severe accidents. SONATA-IV (Simulation of Naturally Arrested Thermal Attack in Vessel) program is set up for in-vessel cooling and a progression of the verification experiment has been done, and an integral verification experiment of the containment integrity for ex-vessel cooling is planned to be carried out based on the separate effect experiments performed in the first phase. First phase study of SONATA-IV is proof of principle experiment and it is composed of LALA (Lower-plenum Arrested Vessel Attack) experiment to find the gap between melt and the lower plenum during melt relocation and to certify melt quenching and CHFG (Critical Heat Flux in Gap) experiment to certify heat transfer mechanism in an artificial gap. As separate effect experiments for ex-vessel cooling, high pressure melt ejection experiment related to the initial condition for debris layer formation in the reactor cavity, crust formation and heat transfer experiment in the molten pool and molten core concrete interaction experiment are performed. (author). 150 refs., 24 tabs., 127 figs.

Antineutrophil cytoplasmic autoantibody-associated small-vessel vasculitis

NARCIS (Netherlands)

Kallenberg, Cees G. M.

Purpose of reviews This review focuses on recent advance in the diagnosis pathogenesis and treatment of antineutrophil cytoplasmic autoantibody-associated small-vessel vasculitis. Recent findings Antineutrophil cytoplasmic autoantibodies are closely associated with Wegener's granulomatosis and

Control of the ORR-PSF pressure-vessel surveillance irradiation experiment temperature

International Nuclear Information System (INIS)

Miller, L.F.

1982-01-01

Control of the Oak Ridge Research Reactor Pool Side Facility (ORR-PSF) pressure vessel surveillance irradiation experiment temperature is implemented by digital computer control of electrical heaters under fixed cooling conditions. Cooling is accomplished with continuous flows of water in pipes between specimen sets and of helium-neon gas in the specimen set housings. Control laws are obtained from solutions of the discrete-time Riccati equation and are implemented with direct digital control of solid state relays in the electrical heater circuit. Power dissipated by the heaters is determined by variac settings and the percent of time that the solid state relays allow power to be supplied to the heaters. Control demands are updated every forty seconds

Characterization of erosion dust and tritiated products inside the jet vessel after the first tritium experiment

International Nuclear Information System (INIS)

Charuau, J.; Belot, Y.; Cetier, P.; Drezet, L.; Grivaud, L.

1992-01-01

An experiment was carried out to characterize the erosion products found in the JET vessel after the first tritium experiment. These products were analyzed for carbon, beryllium, Inconel metals and tritium. All these elements were present in airborne particles or deposited dust. The tritium was found as tritiated water vapour, and also strongly associated to the suspended or deposited particles. It was more abundant in fine than in coarse particles. The particulate tritium seems to be almost entirely 'insoluble' in a water solution

Recent experiences and problems in conducting pressure vessel surveillance examinations

International Nuclear Information System (INIS)

Perrin, J.S.

1979-01-01

Each of the commercial power reactors in the U.S.A. has a pressure vessel surveillance program. The purpose of the programs is to monitor the effects of radiation on the mechanical properties on the steel pressure vessels. A program for a given reactor includes a series of irradiation capsules containing neutron dosimeters and mechanical property specimens. The capsules are periodically removed during the life of the reactor and evaluated. The surveillance capsule examinations conducted to date have been valuable in assessing the effects of radiation on pressure vessels. However, a number of problems have been observed in the course of capsule examinations which potentially could reduce the maximum value of the data obtained. These problems are related to specimen design and preparation, capsule design and preparation, capsule installation and removal, capsule disassembly, specimen testing and evaluation, program documentation, and quality assurance. Examples of problems encountered in the preceding areas are presented in the present paper, and recommendations are made for minimization or prevention of these problems in future programs. Included in the recommendations is that appropriate ASTM standards, ASME Boiler and Pressure Vessel Code sections, and NRC regulations provide the appropriate framework for prevention of problems

Air-water mixing experiments for direct vessel injection of KNGR

International Nuclear Information System (INIS)

Hwang, Do Hyun

2000-02-01

Two air-water mixing experiments are conducted to understand the flow behavior in the downcomer for Direct Vessel Injection (DVI) of Korean Next Generation Reactor (KNGR). In the first experiment which is an air-water experiment in the rectangular channel with the gap size of 1cm, the width of water film is proportional to the water and air velocities and the inclined angle is proportional to the water velocity only, regardless of the water velocity injected in the rectangular channel. It is observed that the amount of entrained water is negligible. In the second experiment which is a full-scaled water jetting experiment without air flow, the width of water film is proportional to the flow rate injected from the pipe exit and the film thickness of water varies from 1.0mm to 5.0mm, and the maximum thickness does not exceed 5.0mm. The amount of water separated from the liquid film after striking of water jetting on the wall is measured. The amount of separation water is proportional to the flow rate, but the separation ratio in the full-scaled water jetting is not over 15%. A simplified physical model, which is designed to predict the trajectories of the width of water film, is validated through the comparison with experiment results. The 13 .deg. upward water droplet of the water injected from the pipe constitutes the outermost boundary at 1.7m below from pipe level, after the water impinges against the wall. In the model, the parameter, η which represents the relationship between the jetting velocity and the initial spreading velocity, is inversely proportional to the water velocity when it impinges against the wall. The error of the predictions by the model is decreased within 14% to the experimental data through use of exponential fitting of η for the jetting water velocity

Safety valve opening and closing operation monitor

International Nuclear Information System (INIS)

Kodama, Kunio; Takeshima, Ikuo; Takahashi, Kiyokazu.

1981-01-01

Purpose: To enable the detection of the closing of a safety valve when the internal pressure in a BWR type reactor is a value which will close the safety valve, by inputting signals from a pressure detecting device mounted directly at a reactor vessel and a safety valve discharge pressure detecting device to an AND logic circuit. Constitution: A safety valve monitor is formed of a pressure switch mounted at a reactor pressure vessel, a pressure switch mounted at the exhaust pipe of the escape safety valve and a logic circuit and the lide. When the input pressure of the safety valve is raised so that the valve and the pressure switch mounted at the exhaust pipe are operated, an alarm is indicated, and the operation of the pressure switch mounted at a pressure vessel is eliminated. If the safety valve is not reclosed when the vessel pressure is decreased lower than the pressure at which it is to be reclosed after the safety valve is operated, an alarm is generated by the logic circuit since both the pressure switches are operated. (Sekiya, K.)

Copper corrosion experiments under anoxic conditions

Energy Technology Data Exchange (ETDEWEB)

Ollila, Kaija [VTT Technical Research Centre of Finland, Espoo (Finland)

2013-06-15

This report gives results from the corrosion experiments with copper under anoxic conditions. The objective was to study whether hydrogen-evolving corrosion reaction could occur. Copper foil samples were exposed in deaerated deionized water in Erlenmeyer flasks in the glove box with inert atmosphere. Four corrosion experiments (Cu1, Cu2, Cu3 and Cu4) were started, as well as a reference test standing in air. Cu1 and Cu2 had gas tight seals, whereas Cu3 and Cu4 had palladium foils as hydrogen permeable enclosure. The test vessels were stored during the experiments in a closed stainless steel vessel to protect them from the trace oxygen of the gas atmosphere and light. After the reaction time of three and a half years, there were no visible changes in the copper surfaces in any of the tests in the glove box, in contrast the Cu surfaces looked shiny and unaltered. The Cu3 test was terminated after the reaction time of 746 days. The analysis of the Pd-membrane showed the presence of H2 in the test system. If the measured amount of 7.2{center_dot}10{sup 5} mol H{sub 2} was the result of formation of Cu{sub 2}O this would correspond to a 200 nm thick corrosion layer. This was not in agreement with the measured layer thickness with SIMS, which was 6{+-}1 nm. A clear weight loss observed for the Cu3 test vessel throughout the test period suggests the evaporation of water through the epoxy sealing to the closed steel vessel. If this occurred, the anaerobic corrosion of steel surface in humid oxygen-free atmosphere could be a source of hydrogen. A similar weight loss was not observed for the parallel test (Cu4). The reference test standing in air showed visible development of corrosion products.

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.

Analysis and optimization on in-vessel inspection robotic system for EAST

Energy Technology Data Exchange (ETDEWEB)

Zhang, Weijun, E-mail: zhangweijun@sjtu.edu.cn; Zhou, Zeyu; Yuan, Jianjun; Du, Liang; Mao, Ziming

2015-12-15

Since China has successfully built her first Experimental Advanced Superconducting TOKAMAK (EAST) several years ago, great interest and demand have been increasing in robotic in-vessel inspection/operation systems, by which an observation of in-vessel physical phenomenon, collection of visual information, 3D mapping and localization, even maintenance are to be possible. However, it has been raising many challenges to implement a practical and robust robotic system, due to a lot of complex constraints and expectations, e.g., high remanent working temperature (100 °C) and vacuum (10{sup −3} pa) environment even in the rest interval between plasma discharge experiments, close-up and precise inspection, operation efficiency, besides a general kinematic requirement of D shape irregular vessel. In this paper we propose an upgraded robotic system with redundant degrees of freedom (DOF) manipulator combined with a binocular vision system at the tip and a virtual reality system. A comprehensive comparison and discussion are given on the necessity and main function of the binocular vision system, path planning for inspection, fast localization, inspection efficiency and success rate in time, optimization of kinematic configuration, and the possibility of underactuated mechanism. A detailed design, implementation, and experiments of the binocular vision system together with the recent development progress of the whole robotic system are reported in the later part of the paper, while, future work and expectation are described in the end.

An interior vessel viewing system for DIII-D

International Nuclear Information System (INIS)

Senior, R.

1989-11-01

It was anticipated that there could be damage to the interior walls of the vacuum vessel during operations of the DIII-D tokamak. A method of viewing the inside of the vessel from the outside was required, that would allow the interior walls to be inspected visually for damage and to locate any debris resulting from operations. A miniature closed circuit television color camera system was developed which could be inserted into one of several ports of the vessel during a 'clean' vent, i.e., vented to inert gas. The system has pan, tilt and zoom capability and carries its own lighting. The use of this system allows a quick assessment of the condition of the vessel to be made under 'clean' vent conditions. This precludes the need for the permit process and manned entry into the vessel which would allow air inside the vessel. A permanent record of the inspection can then be made on video tape. The design and configuration of this camera system is presented and its use as a diagnostic tool discussed. 2 refs., 5 figs

Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

International Nuclear Information System (INIS)

Zado, V.

1998-01-01

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

Warm pre-stress experiments on highly irradiated reactor pressure vessel steel

International Nuclear Information System (INIS)

Landron, C.; Ait-Bachir, M.; Moinereau, D.; Molinie, E.; Garbay, E.

2015-01-01

In the aim to justify in-service integrity of reactor pressure vessel beyond 40 years, experimental warm pre-stress (WPS) tests were performed on irradiated materials representative of RPV steels corresponding to 40 operating years. Different types of WPS loading path have been considered to cover typical postulated accidental transients. These results confirmed the beneficial effect of WPS on the cleavage fracture resistance of the irradiated materials. No fracture occurred during the cooling phase of the loading path and the fracture toughness values are higher than that measured with conventional isothermal tests. The analyses of the experiments, conducted using either simplified engineering models or more refined fracture models based on local approach to cleavage fracture, are in agreement with the experimental results. (authors)

An experimental study on feasibility of ex-vessel cooling through the external guide vessel

International Nuclear Information System (INIS)

Kang, Kyoung-Ho; Kim, Jong-Hwan; Park, Rae-Jun; Kim, Sang-Baik

2000-01-01

This paper presents the results of a series of experiments for assessing the efficacy of ex-vessel cooling through the external guide vessel during a severe accident. Four tests were performed in the LAVA test facility at KAERI, varying the boundary conditions at the outer surface of the vessel. The first test was a dry condition test conducted without cooling the outside of the vessel. On the other hand, in the second test, the cooling of the vessel surface was produced by gravity-driven forced injection of water along the annular gap of 25 mm between the vessel and the external guide vessel. Water flow rate was about 0.85 kg/s and total mass of available water was 300 kg. For the evaluation of the water flow rate effect, the third test was performed with a pool type cooling in the annulus without any circulation of water. These two external cooling tests were performed under elevated pressure of about 1.6 MPa. Finally, the fourth test was conducted under atmospheric pressure to evaluate the effect of system pressure on boiling heat transfer characteristics. In the dry test and the pool type ex-vessel cooling test performed under atmospheric pressure, the vessel was failed by a melt penetration at about 40 degree upper position from the vessel bottom, which is coincident with the boundary of the Al 2 O 3 /Fe melt separated layers. On the other hand, in both of the ex-vessel cooling tests conducted under elevated pressure of about 1.6 MPa, the vessel didn't fail. Compared with the pool boiling test, the vessel experienced effective cooling due to the inlet flow in the forced flow test. Synthesized the results of the tests, it was shown that the heat removal with ex-vessel cooling through the guide vessel is feasible, but the additional evaluations should be performed to guarantee enough thermal margin. (author)

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.

PWR vessel inspection performance improvements

International Nuclear Information System (INIS)

Blair Fairbrother, D.; Bodson, Francis

1998-01-01

A compact robot for ultrasonic inspection of reactor vessels has been developed that reduces setup logistics and schedule time for mandatory code inspections. Rather than installing a large structure to access the entire weld inspection area from its flange attachment, the compact robot examines welds in overlapping patches from a suction cup anchor to the shell wall. The compact robot size allows two robots to be operated in the vessel simultaneously. This significantly reduces the time required to complete the inspection. Experience to date indicates that time for vessel examinations can be reduced to fewer than four days. (author)

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

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.

Vessel used in radiation counting to determine radioactivity levels

International Nuclear Information System (INIS)

Charlton, J.C.; Glover, J.S.; Shephard, B.P.

1977-01-01

This invention concerns the vessels used in radiation counting to determine radioactivity levels. These vessels prove to be particularly useful in analyses of the kind where a radioactive element or compound is separated into two phases and the radioactivity of one phase is determined. Such a vessel used in the counting of radiation includes an organic plastic substance tube appreciably cylindrical in shape whose upper end is open whilst the lower end is closed and integral with it, and an anti-radiation shield in metal or in metal reinforced plastic located at the lower end of the tube and extending along the wall of the tube up to a given height. The vessel contains a reaction area of 1 to 10 ml for holding fluid reagents [fr

Experiment on transient heat transfer in closed narrow channel

International Nuclear Information System (INIS)

Ochiai, Masaaki

1985-01-01

Heat transfer coefficients and transient pressures in closed narrow channels were obtained experimentally, in order to assess the gap heat transfer models in the computer code WTRLGD which were devised to analyze the internal pressure behavior of waterlogged fuel rods. Gap widths of channels are 0.1--0.5mm to simulate the gap region of waterlogged fuel rods, and test fluids are water (7--89.2 0 C) and Freon-113 (9.2 0 C). The results show that the heater temperature and the pressure measured in the experiments without the DNB occurrence are simulated fairly well by the calculational model of WTRLGD where the heat transfer in a closed narrow channel is evaluated with one-dimensional transient thermal conduction equation and Jens and Lottes' correlation for nucleate boiling. Consequently, it is also suggested that the above equations are available for evaluation of heat flux from fuel to internal water of waterlogged fuel rods. The film boiling heat transfer coefficient was in the same order of that evaluated by Bromley's correlation and the DNB heat flux was smaller than that obtained in quasi-steady experiments with ordinary systems, although the experimental data for them were not enough. (author)

Feedback from Westinghouse experience on segmentation of reactor vessel internals - 59013

International Nuclear Information System (INIS)

Kreitman, Paul J.; Boucau, Joseph; Segerud, Per; Fallstroem, Stefan

2012-01-01

With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive water-jet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR's), Boiling Water Reactors (BWR's), Gas Cooled Reactors (GCR's) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a

Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 4. Numerical analysis of 1/10 scaled water experiment with the AQUA code

International Nuclear Information System (INIS)

Muramatu, Toshiharu; Yamaguchi, Akira

2004-01-01

A large-scale sodium-cooled fast breeder reactor in the feasibility studies on commercialized fast reactors has a feature of consideration of thorough simplified and compacted systems and components design to realize drastic economical improvements. Therefore, special attentions should be paid to thermohydraulic designs for gas entrainment behavior from free surface, flow-induced vibration of in-vessel components, thermal stratification in the plenum, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. A numerical analysis was carried out with a multi-dimensional code AQUA to confirm an applicability to the evaluations for the in-vessel thermohydraulic phenomena using a 1/10 scaled water experiment simulating the large-scale fast breeder reactor in the feasibility studies. From the analysis, the following results were obtained. (1) In-vessel thermohydraulics characterized by a radiated flow pattern to the reactor vessel wall and a strong upward flow through a slit of the upper core structures were evaluated. These characteristics agreed approximately with the water experiment. (2) The upward velocity values at the slit agreed well with the experimental data under a condition of γ z = 0.3 and ξ z = 0.5, though overall evaluations of the in-vessel thermohydraulics were failed to predict quantitatively. (3) The AQUA code is applicable to the in-vessel thermohydraulics evaluations in the feasibility studies, though it is necessary to make further modifications of the calculational models for accurate evaluations. On the one hand, it was confirmed that calculated results for the 1/10 water experimental model and the 1/1 actual-scaled model agreed quantitatively for the in-vessel thermohydraulics characteristics indicated above. (author)

Nuclear reactor installation with outer shell enclosing a primary pressure vessel

International Nuclear Information System (INIS)

1975-01-01

The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel containing the heat source, an outer shell enclosing the primary pressure vessel and acting as a secondary means of containment for this vessel against outside projectiles. Multiple auxiliary equipment points are arranged outside the outer shell which comprises a part of a lower wall around the primary pressure vessel, an annular part integrated in the lower wall and extending outwards as from this wall and an upper part integrated in the annular part and extending above this annular part and above the primary pressure vessel. The annular part and the primary pressure vessel are formed with vertical penetrations which can be closed communicating respectively with the auxiliary equipment points and with inside the pressure vessel whilst handling gear is provided in the upper part for vertically raising reactor components through these penetrations and for transporting them over the annular part and over the primary pressure vessel [fr

Containment vessel construction for nuclear power reactors

International Nuclear Information System (INIS)

Sulzer, H.D.; Coletti, J.L.

1975-01-01

A nuclear containment vessel houses an inner reactor housing structure whose outer wall is closely spaced from the inner wall of the containment vessel. The inner reactor housing structure is divided by an intermediate floor providing an upper chamber for housing the reactor and associated steam generators and a lower chamber directly therebeneath containing a pressure suppression pool. Communication between the upper chamber and the pressure suppression pool is established by conduits extending through the intermediate floor which terminate beneath the level of the pressure suppression pool and by inlet openings in the reactor housing wall beneath the level of the pressure suppression pool which communicate with the annulus formed between the outer wall of the reactor housing structure and the inner wall of the containment vessel. (Official Gazette)

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

An investigation of the flow dependence of temperature gradients near large vessels during steady state and transient tissue heating

International Nuclear Information System (INIS)

Kolios, M.C.; Worthington, A.E.; Hunt, J.W.; Holdsworth, D.W.; Sherar, M.D.

1999-01-01

Temperature distributions measured during thermal therapy are a major prognostic factor of the efficacy and success of the procedure. Thermal models are used to predict the temperature elevation of tissues during heating. Theoretical work has shown that blood flow through large blood vessels plays an important role in determining temperature profiles of heated tissues. In this paper, an experimental investigation of the effects of large vessels on the temperature distribution of heated tissue is performed. The blood flow dependence of steady state and transient temperature profiles created by a cylindrical conductive heat source and an ultrasound transducer were examined using a fixed porcine kidney as a flow model. In the transient experiments, a 20 s pulse of hot water, 30 deg. C above ambient, heated the tissues. Temperatures were measured at selected locations in steps of 0.1 mm. It was observed that vessels could either heat or cool tissues depending on the orientation of the vascular geometry with respect to the heat source and that these effects are a function of flow rate through the vessels. Temperature gradients of 6 deg. C mm -1 close to large vessels were routinely measured. Furthermore, it was observed that the temperature gradients caused by large vessels depended on whether the heating source was highly localized (i.e. a hot needle) or more distributed (i.e. external ultrasound). The gradients measured near large vessels during localized heating were between two and three times greater than the gradients measured during ultrasound heating at the same location, for comparable flows. Moreover, these gradients were more sensitive to flow variations for the localized needle heating. X-ray computed tomography data of the kidney vasculature were in good spatial agreement with the locations of all of the temperature variations measured. The three-dimensional vessel path observed could account for the complex features of the temperature profiles. The flow

Non-contact method of search and analysis of pulsating vessels

Science.gov (United States)

Avtomonov, Yuri N.; Tsoy, Maria O.; Postnov, Dmitry E.

2018-04-01

Despite the variety of existing methods of recording the human pulse and a solid history of their development, there is still considerable interest in this topic. The development of new non-contact methods, based on advanced image processing, caused a new wave of interest in this issue. We present a simple but quite effective method for analyzing the mechanical pulsations of blood vessels lying close to the surface of the skin. Our technique is a modification of imaging (or remote) photoplethysmography (i-PPG). We supplemented this method with the addition of a laser light source, which made it possible to use other methods of searching for the proposed pulsation zone. During the testing of the method, several series of experiments were carried out with both artificial oscillating objects as well as with the target signal source (human wrist). The obtained results show that our method allows correct interpretation of complex data. To summarize, we proposed and tested an alternative method for the search and analysis of pulsating vessels.

Pressure vessel lid

International Nuclear Information System (INIS)

Schoening, J.; Elter, C.; Becker, G.; Pertiller, S.

1986-01-01

The invention concerns a lid for closing openings in reactor pressure vessels containing helium, which is made as a circular casting with hollow spaces and a flat floor and is set on the opening and kept down. It consists of helium-tight metal cast material with sufficient temperature resistance. There are at least two concentric heat resistant seals let into the bottom of the lid. The bottom is in immediate contact with the container atmosphere and has hollow spaces in its inside in the area opposite to the opening. (orig./HP) [de

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.

Eyes wide shut: amygdala mediates eyes-closed effect on emotional experience with music.

Science.gov (United States)

Lerner, Yulia; Papo, David; Zhdanov, Andrey; Belozersky, Libi; Hendler, Talma

2009-07-15

The perceived emotional value of stimuli and, as a consequence the subjective emotional experience with them, can be affected by context-dependent styles of processing. Therefore, the investigation of the neural correlates of emotional experience requires accounting for such a variable, a matter of an experimental challenge. Closing the eyes affects the style of attending to auditory stimuli by modifying the perceptual relationship with the environment without changing the stimulus itself. In the current study, we used fMRI to characterize the neural mediators of such modification on the experience of emotionality in music. We assumed that closed eyes position will reveal interplay between different levels of neural processing of emotions. More specifically, we focused on the amygdala as a central node of the limbic system and on its co-activation with the Locus Ceruleus (LC) and Ventral Prefrontal Cortex (VPFC); regions involved in processing of, respectively, 'low', visceral-, and 'high', cognitive-related, values of emotional stimuli. Fifteen healthy subjects listened to negative and neutral music excerpts with eyes closed or open. As expected, behavioral results showed that closing the eyes while listening to emotional music resulted in enhanced rating of emotionality, specifically of negative music. In correspondence, fMRI results showed greater activation in the amygdala when subjects listened to the emotional music with eyes closed relative to eyes open. More so, by using voxel-based correlation and a dynamic causal model analyses we demonstrated that increased amygdala activation to negative music with eyes closed led to increased activations in the LC and VPFC. This finding supports a system-based model of perceived emotionality in which the amygdala has a central role in mediating the effect of context-based processing style by recruiting neural operations involved in both visceral (i.e. 'low') and cognitive (i.e. 'high') related processes of emotions.

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

Spent nuclear fuel assembly storage vessel

International Nuclear Information System (INIS)

Yagishita, Takuya

1998-01-01

The vessel of the present invention promotes an effect of removing after heat of spent nuclear fuel assemblies so as not to give force to the storage vessel caused by expansion of heat removing partitioning plates. Namely, the vessel of the present invention comprises a cylinder body having closed upper and lower portions and a plurality of heat removing partitioning cylinders disposed each at a predetermined interval in the circumferential direction of the above-mentioned cylinder body. The heat removing partitioning cylinders comprises (1) first heat removing partitioning plates extended in the radial direction of the cylinder body and opposed at a predetermined gap in the circumferential direction of the cylinder body, and having the base ends on the side of the inner wall of the cylinder body being secured to the inner wall of the cylinder body and (2) a second heat removing plate for connecting the top ends of both opposed heat removing partitioning plates on the central side of the cylinder body with each other. Spent nuclear fuel assemblies are contained in a plurality of closed spaces surrounded by the first heat removing partitioning plates and the second heat removing partitioning plate. With such constitution, since after heat is partially transferred from the heat removing partitioning plates to the cylindrical body directly by heat conduction, the heat removing effect can be promoted compared with the prior art. (I.S.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Light-induced bird strikes on vessels in Southwest Greenland

DEFF Research Database (Denmark)

Merkel, Flemming Ravn; Johansen, Kasper Lambert

2011-01-01

Light-induced bird strikes are known to occur when vessels navigate during darkness in icy waters using powerful searchlight. In Southwest Greenland, which is important internationally for wintering seabirds, we collected reports of incidents of bird strikes over 2–3 winters (2006–2009) from navy...... vessels, cargo vessels and trawlers (total n = 19). Forty-one incidents were reported: mainly close to land (birds were reported killed in a single incident. All occurred between 5 p.m. and 6 a.m. and significantly more birds were involved when...... visibility was poor (snow) rather than moderate or good. Among five seabird species reported, the common eider (Somateria mollissima) accounted for 95% of the bird casualties. Based on spatial analyses of data on vessel traffic intensity and common eider density we are able to predict areas with high risk...

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.

Experimental study on in-vessel debris coolability during severe accident

International Nuclear Information System (INIS)

Kim, S. B.; Park, R. J.; Kim, H. D.

2002-05-01

A research program, called SONATA-IV(Simulation of Naturally Arrested Thermal Attack In-Vessel), has been performed to verify the gap cooling mechanism of corium in the lower plenum, and to develop management and mitigation strategies under severe accident conditions. For the proof-of-principles experiment, the LAVA(Lower-plenum Arrested Vessel Attack) experiments have been performed to gather proof of gap formation and to evaluate the gap effect on in-vessel cooling, using Al 2 O 3 /Fe (or Al 2 O 3 only) thermite melt as corium simulant. And also the CHFG(Critical Heat Flux in Gap) experiments have been performed to measure the critical power and to investigate the inherent cooling mechanism in the hemispherical narrow gap. In addition to the experiments, LILAC code was developed to analyze and predict the thermo-hydraulic phenomena of the corium relocated in the reactor lower plenum. It could be found from the LAVA and CHFG experimental results that continuous gap ranged from 1 to 5 mm was formed and that maximum heat removal capacity through a gap is a key factor in determining the potentials of the integrity of the vessel. After all the possibility of IVR(In-Vessel corium Retention) through gap cooling highly depends on the melt relocated into the lower plenum and the gap size. So, feasibility experiments have been performed for the assessment of improved IVR concepts using an internal engineered gap device and a dual strategy of In/Ex-vessel cooling using the LAVA facility. It is preliminarily concluded that these cooling measures lead to an enhanced cooling of the corium in the lower plenum of the reactor vessel. The additional studies will be performed to verify the quantitative heat removal capacity for these cooling measures in the 2nd phase of mid- and long term project period

Slideline verification for multilayer pressure vessel and piping analysis

International Nuclear Information System (INIS)

Van Gulick, L.A.

1983-01-01

Nonlinear finite element method (FEM) computer codes with slideline algorithm implementations should be useful for the analysis of prestressed multilayer pressure vessels and piping. This paper presents closed form solutions useful for validating slideline implementations for this purpose. The solutions describe stresses and displacements of an internally pressurized elastic-plastic sphere initially separated from an elastic outer sphere by a uniform gap. Comparison of closed form and FEM results evaluates the usefulness of the closed form solution and the validity of the slideline implementation used

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

Pressure test method for reactor pressure vessel in construction field

International Nuclear Information System (INIS)

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

1998-01-01

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

36 CFR 13.1188 - Where to get charts depicting closed waters.

Science.gov (United States)

2010-07-01

... closed waters. 13.1188 Section 13.1188 Parks, Forests, and Public Property NATIONAL PARK SERVICE... and Preserve Vessel Operating Restrictions § 13.1188 Where to get charts depicting closed waters. Closed waters and islands within Glacier Bay as described in §§ 13.1174-13.1180 of this subpart are...

Heat load imposed on reactor vessels during in-vessel retention of core melts

Energy Technology Data Exchange (ETDEWEB)

Kim, Su-Hyeon; Chung, Bum-Jin, E-mail: bjchung@khu.ac.kr

2016-11-15

Highlights: • Angular heat load on reactor vessel by natural convection of oxide pool was measured. • High Ra was achieved by using mass transfer experiments based on analogy concept. • Measured Nusselt numbers agreed reasonably with the other existing studies. • Three different types of volumetric heat sources were compared. • They didn’t affect the heat flux of the top plate but affected those of the reactor vessel. - Abstract: We measured the heat load imposed on reactor vessels by natural convection of the oxide pool in severe accidents. Based on the analogy between heat and mass transfer, mass transfer experiments were performed using a copper sulfate electroplating system. A modified Rayleigh number of the order 10{sup 14} was achieved in a small facility with a height of 0.1 m. Three different types of volumetric heat sources were compared and the average Nusselt number of the curved surface was 39% lower, whereas in the case of the top plate was 6% higher than in previous studies with a two-dimensional geometry due to the high Sc value of this study. Reliable experimental data on the angular heat flux ratios were reported compared to those of the BALI and SIGMA CP facilities in terms of fluctuations and consistency.

Behaviour of Viscoelastic - Viscoplastic Spheres and Cylinders - Fully Plastic Vessel Walls

DEFF Research Database (Denmark)

Ottosen, N. Saabye

1985-01-01

The material model consists of a viscoelastic Burgers element and an additional viscoplastic Bingham element when the effective stress exceeds the yield stress. For fully plastic vessel walls, exact closed-form expressions arc derived for the stress and strain state in pressurised or relaxation...... loaded thick-walled cylinders in plane strain and spheres. For the spherical problem, the material compressibility is accounted for. The influence of the different material parameters on the behaviour of the vessels is evaluated. It is shown that the magnitude of the Maxwell viscosity is of major...... importance for the long-term behaviour of thick-walled fully plastic vessels....

Pressure tube rupture in a closed tank

International Nuclear Information System (INIS)

Khater, H.A.; Hadaller, G.I.; Stern, F.

1985-06-01

A study has been prepared on the feasibility of conducting pressure tube/calandria tube rupture tests in a closed tank, simulating a scaled-down calandria vessel. The study includes: i) a review of previous work, ii) an analytical investigation of the scaling problem of the calandria vessel and relevant in-core structures, iii) selection of a method for initiating pressure tube/calandria tube rupture, iv) a set of specifications for the test assembly, v) general arrangement drawings, vi) a proposal for a test matrix, vii) a survey and evaluation of existing facilities which could provide the required high pressure, temperature and fluid inventory, and viii) a cost estimate for the detailed design and construction, instrumentation, data acquisition and reduction, testing and reporting. The study concludes that it is both technically and practically feasible to conduct pressure tube rupture tests in a closed tank

Pressure vessel integrity 1991

International Nuclear Information System (INIS)

Bhandari, S.; Doney, R.O.; McDonald, M.S.; Jones, D.P.; Wilson, W.K.; Pennell, W.E.

1991-01-01

This volume contains papers relating to the structural integrity assessment of pressure vessels and piping, with special emphasis on nuclear industry applications. The papers were prepared for technical sessions developed under the sponsorship of the ASME Pressure Vessels and Piping Division Committees for Codes and Standards, Computer Technology, Design and Analysis, and Materials Fabrication. They were presented at the 1991 Pressure Vessels and Piping Division Conference in San Diego, California, June 23-27. The primary objective of the sponsoring organization is to provide a forum for the dissemination and discussion of information on development and application of technology for the structural integrity assessment of pressure vessels and piping. This publication includes contributions from authors from Australia, France, Japan, Sweden, Switzerland, the United Kingdom, and the United States. The papers here are organized in six sections, each with a particular emphasis as indicated in the following section titles: Fracture Technology Status and Application Experience; Crack Initiation, Propagation and Arrest; Ductile Tearing; Constraint, Stress State, and Local-Brittle-Zones Effects; Computational Techniques for Fracture and Corrosion Fatigue; and Codes and Standards for Fatigue, Fracture and Erosion/Corrosion

Closure system of a vessel made of prestressed concrete

International Nuclear Information System (INIS)

Audibert, Alain

1974-01-01

The present invention relates to removable plugs of prestressed concrete which can be fitted to every type of closed high pressure vessels and especially to the cylindrical vessels of nuclear reactors. The method involved permits the plug to be fitted to the vessel through both radial and axial prestress. In this purpose, said invention proposes removable prestress ribs fitted inside sheaths in the plug and extending throughout the upper part of the bearing surfaces of the plug, said ribs being regularly arranged along the generators of an hyperboloid of one sheet. Owing to this important feature, that is to say said inclination of the ribs in accordance with the generators of said hyperboloid, said rib inclination can be changed on requirement for each realization [fr

Feasibility evaluation of 3D photoacoustic imaging of blood vessel structure using multiple wavelengths with a handheld probe

Science.gov (United States)

Uchimoto, Yo; Namita, Takeshi; Kondo, Kengo; Yamakawa, Makoto; Shiina, Tsuyoshi

2018-02-01

Photoacoustic imaging is anticipated for use in portraying blood vessel structures (e.g. neovascularization in inflamed regions). To reduce invasiveness and enhance ease handling, we developed a handheld photoacoustic imaging system using multiple wavelengths. The usefulness of the proposed system was investigated in phantom experiments and in vivo measurements. A silicon tube was embedded into chicken breast meat to simulate the blood vessel. The tube was filled with ovine blood. Then laser light was guided to the phantom surface by an optical fiber bundle close to the linear ultrasound probe. Photoacoustic images were obtained at 750-950 nm wavelengths. Strong photoacoustic signals from the boundary between blood and silicon tube are observed in these images. The shape of photoacoustic spectrum at the boundary resembles that of the HbO2 absorption spectrum at 750-920 nm. In photoacoustic images, similarity between photoacoustic spectrum and HbO2 absorption spectrum was evaluated by calculating the normalized correlation coefficient. Results show high correlation in regions of strong photoacoustic signals in photoacoustic images. These analyses demonstrate the feasibility of portraying blood vessel structures under practical conditions. To evaluate the feasibility of three-dimensional vascular imaging, in vivo experiments were conducted using three wavelengths. A right hand and ultrasound probe were set in degassed water. By scanning a probe, cross-sectional ultrasound and photoacoustic images were obtained at each location. Then, all ultrasound or photoacoustic images were piled up respectively. Then three-dimensional images were constructed. Resultant images portrayed blood vessel-like structures three-dimensionally. Furthermore, to distinguish blood vessels from other tissues (e.g. skin), distinguishing images of them were constructed by comparing photoacoustic signal intensity among three wavelengths. The resultant image portrayed blood vessels as

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

A novel methodology for adaptive wave filtering of marine vessels: Theory and experiments

Digital Repository Service at National Institute of Oceanography (India)

Hassani, V.; Pascoal, A.M.; Sorensen, A.J.

This paper addresses a filtering problem that arises in the design of dynamic positioning systems for ships and offshore rigs subjected to the influence of sea waves. The vessel`s dynamic model adopted captures the sea state as an uncertain...

Stresses in reactor pressure vessel nozzles -- Calculations and experiments

International Nuclear Information System (INIS)

Brumovsky, M.; Polachova, H.

1995-01-01

Reactor pressure vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a reactor pressure vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber's, Hardrath-Ohman's as well as equivalent energy ones, used in different reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared

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.)

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.

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.

Westinghouse experience in using mechanical cutting for reactor vessel internals segmentation

International Nuclear Information System (INIS)

Boucau, Joseph; Fallstroem, Stefan; Segerud, Per; Kreitman, Paul J.

2010-01-01

plants dismantled to date in the US have repackaged the less activated waste back into the reactor vessel and shipped the entire assembly to the disposal site. Decisions like these can be driven by many factors such as disposal costs, transportation logistics, licensing fees, etc., but will have a significant impact on the segmentation and packaging plan so must be considered early in the planning phase. All segmentation tools are remotely controlled since the mechanical segmentation projects that Westinghouse has executed, so far, have been performed under water due to the high radiation levels. ALARA and personal safety is the number one priority during the site work. The complexity of the work requires well designed and reliable tools. Westinghouse has optimized the technologies from its experiences accumulated over the years. Its main focus has always been to improve tool handling and cutting speed, water cleanliness, fail-safe and safety aspects. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. The purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date. (authors)

A system for the thermal insulation of a pre-stressed concrete vessel

International Nuclear Information System (INIS)

Aubert, Gilles; Petit, Guy.

1975-01-01

This invention concerns the thermal insulation of a pre-stressed concrete vessel for a pressurised water nuclear reactor, this vessel being fitted internally with a leak-proof metal lining. Two rings are placed at the lower and upper parts of the vessel respectively. The upper ring is closed with a cover. These rings differ in diameter, are fitted with a metal insulating and mark the limits of a chamber between the vaporisable fluid and the internal wall of the vessel. This chamber is filled with a fluid in the liquid phase up to the liquid/vapor interface level of the fluid and with a gas above that level, the covering of the rings forming a cold fluid liquid seal. Each ring is supported by the vessel. Leak-proof components take up the radial expansion of the rings [fr

Design and fabrication of the vacuum vessel for the Advanced Toroidal Facility

International Nuclear Information System (INIS)

Chipley, K.K.; Frey, G.N.

1985-01-01

The vacuum vessel for the Advanced Toroidal Facility (ATF) is a heavily contoured and very complex formed vessel that is specifically designed to allow for maximum plasma volume in a pure stellarator arrangement. The design of the facility incorporates an internal vessel that is closely fitted to the two helical field coils following the winding law theta = 1/6phi. Metallic seals have been incorporated throughout the system to minimize impurities. The vessel has been fabricated utilizing a comprehensive set of tooling fixtures specifically designed for the task of forming 6-mm stainless steel plate to the complex shape. Computer programs were used to develop a series of ribs that essentially form an internal mold of the vessel. Plates were press-formed with multiple compound curves, fitted to the fixture, and joined with full-penetration welds. 7 refs., 8 figs

Early Experiences with Family Conflict: Implications for Arguments with a Close Friend.

Science.gov (United States)

Herrera, Carla; Dunn, Judy

1997-01-01

Examined associations between children's early experiences in family disputes and later conflict management with close friends. Found that argument used by mothers and siblings that considered children's needs was positively associated with children's later constructive argument and resolution techniques. Mothers' use of argument predicted…

Cold source vessel development for the advanced neutron source

Energy Technology Data Exchange (ETDEWEB)

Williams, P.T.; Lucas, A.T. [Oak Ridge National Lab., TN (United States)

1995-09-01

The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory (ORNL), will be a user-oriented neutron research facility that will produce the most intense flux of neutrons in the world. Among its many scientific applications, the productions of cold neutrons is a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410 mm diameter sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel`s inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design are being performed with multi-dimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This paper presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that will be used to verify the final design.

Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

International Nuclear Information System (INIS)

Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

1984-01-01

The results of reactor material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address exvessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debrids characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity. (orig.)

Forty years of experience on closed-cycle gas turbines

International Nuclear Information System (INIS)

Keller, C.

1978-01-01

Forty years of experience on closed-cycle gas turbines (CCGT) is emphasized to substantiate the claim that this prime-mover technology is well established. European fossil-fired plants with air as the working fluid have been individually operated over 100,000 hours, have demonstrated very high availability and reliability, and have been economically successful. Following the initial success of the small air closed cycle gas turbine plants, the next step was the exploitation of helium as the working fluid for plants above 50 MWe. The first fossil fired combined power and heat plant at Oberhausen, using a helium turbine, plays an important role for future nuclear systems and this is briefly discussed. The combining of an HTGR and an advanced proven power conversion system (CCGT) represents the most interesting and challenging project. The key to acceptance of the CCGT in the near term is the introduction of a small nuclear cogeneration plant (100 to 300 MWe) that utilizes the waste heat, demonstrating a very high fuel utilization efficiency: aspects of such a plant are outlined. (author)

Experience in dismantling and packaging of pressure vessel and core internals

International Nuclear Information System (INIS)

Pillokat, Peter; Bruhn, Jan Hendrik

2011-01-01

Nuclear Company AREVA is proud to look back on versatile experience in successfully dismantling nuclear components. After performing several minor dismantling projects and studies for nuclear power plants, AREVA completed the order for dismantling of all remaining Reactor Pressure Vessel internals at German Boiling Water Reactor Wuergassen NPP in October '08. During the onsite activities about 121 tons of steel were successfully cut and packed under water into 200l- drums, as the dismantling was performed partly in situ and partly in an underwater working tank. AREVA deployed a variety of different cutting techniques such as band sawing, milling, nibbling, compass sawing and water jet cutting throughout this project. After successfully finishing this task, AREVA dismantled the cylindrical part of the Wuergassen Pressure Vessel. During this project approximately 320 tons of steel were cut and packaged for final disposal, as dismantling was mainly performed by on air use of water jet cutting with vacuum suction of abrasive and kerfs material. The main clue during this assignment was the logistic challenge to handle and convey cut pieces from the pressure vessel to the packing area. For this, an elevator was installed to transport cut segments into the turbine hall, where a special housing was built for final storage conditioning. At the beginning of 2007, another complex dismantling project of great importance was acquired by AREVA. The contract included dismantling and conditioning for final storage of the complete RPV Internals of the German Pressurized Water Reactor Stade NPP. Very similar cutting techniques turned out to be the proper policy to cope this task. On-site activities took place in up to 5 separate working areas including areas for post segmentation and packaging to perform optimized parallel activities. All together about 85 tons of Core Internals were successfully dismantled at Stade NPP until September '09. To accomplish the best possible on

Phenomenological vessel burst investigations

International Nuclear Information System (INIS)

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

1985-07-01

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

Benchmark study of shear buckling of a cylindrical vessel

International Nuclear Information System (INIS)

Dostal, M.; Austin, N.M.; Peano, A.; Combescure, A.; Bastien, R.; Carnoy, E.G.

1986-01-01

The possibility of a buckling failure of the primary vessel subjected to seismic excitation has been considered, by all major designers of loop and pool type liquid metal cooled fast breeder reactors. The problem is particularly onerous in this type of reactor due to their large size, coupled with small wall thicknesses. This report details the results of the first phase in a joint European code validation exercise on the static shear buckling behaviour of thin, low aspect ratio stainless steel cylinders. Linear and non-linear finite element analyses were performed by four organizations using three different computer codes, i.e. NNC (UK)-ABAQUS, ISMES (Italy)-ABAQUS, CEA (France)-BILBO/INCA and NOVATOME (France)-NOVNL. The computed results were compared directly with experimental results. It was discovered that refined finite element models were essential if accurate buckling loads were to be calculated. Buckling analyses in 3D were therefore computationally expensive and 2D analyses, where applicable, proved an useful alternative. Traditional linear (Euler) bifurcation analysis seriously over-estimated the buckling loads by around 50 %. Extrapolation techniques can however be used to reduce this discrepancy. Elasto-plastic bifurcation analysis predicted conservative buckling loads close to the experimental value. Non-linear, large displacement analyses were performed on the vessel. The effect of geometrical imperfections in the vessel was considered. These analyses all over-estimated the experimental buckling load by 10 %-25 % and appeared to be largely insensitive to the initial imperfection size. Each of the codes appeared to predict reasonably well the final buckled geometry although the analytical load-deflection estimate did not agree exactly with the experiment

50 CFR 648.81 - NE multispecies closed areas and measures to protect EFH.

Science.gov (United States)

2010-10-01

...) Essential Fish Habitat Closure Areas. (1) In addition to the restrictions under paragraphs (a) through (e... vessel may enter, fish, or be in the area known as Closed Area I (copies of a chart depicting this area... vessels are adversely affecting habitat or NE multispecies stocks, the Regional Administrator may, through...

A quantitative methodology for reactor vessel pressurized thermal shock decision making

International Nuclear Information System (INIS)

Ackerson, D.S.; Balkey, K.R.; Meyer, T.A.; Ofstun, R.P.; Rupprecht, S.D.; Sharp, D.R.

1983-01-01

The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). Previous reactor vessel integrity concerns have led to changes in vessel and plant system design and to operating procedures, and increased attention to the PTS issue is causing consideration of further modifications. Events such as excess feedwater, loss of normal feedwater, and steam generator tube rupture have led to significant primary system cooldowns. Each of these cooldown transients occurred concurrently with a relatively high primary system pressure. Considerations of these and other postulated cooldown events has drawn attention to the impact of operator action and control system effects on reactor vessel PTS. A methodology, which couples event sequence analysis with probabilistic fracture mechanics analyses, was developed to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. (orig./RW)

Contribution of materials investigations and operating experience of reactor vessel internals to PWRs' safety, performance and reliability

International Nuclear Information System (INIS)

Lemaire, E.; Monteil, N.; Jardin, N.; Doll, M.

2015-01-01

The Reactor Pressure Vessel Internals (RVI) include all the components inside the pressure vessel, except the nuclear fuel, the rod cluster assemblies and the instrumentation. The RVI consist of bolted and welded structures that are divided into two sub-assemblies: the upper internals which are removed at every refueling outage and the lower internals which are systematically removed for inspection at every 10-year outage. The main functions of the RVI are to position the core, to support it, and to provide a coolant flow by channeling the fluid. Moreover, the lower internals contribute to a neutron protection of the reactor pressure vessel by absorbing most of the neutron flux from the core. Depending on their location and material composition, the RVI components can face different ageing phenomena, that are actual or potential (such as wear, fatigue, stress corrosion cracking, irradiation assisted stress corrosion cracking, hardening and loss of ductility due to neutron irradiation, irradiation creep and irradiation swelling). EDF has developed a strategy for managing ageing and demonstrating the capacity of the RVI to perform their design functions over 40 years of operation. This overall approach is periodically revisited to take into account the most recent knowledge obtained from the following main topics: Safety Analyses, Research-Development programs, In-Service Inspection (ISI) results, Maintenance programs and Metallurgical Examinations. Based on continuous improvements in those fields, the goal of this paper is to present the way that materials investigations and operating experience obtained on RVI are managed by EDF to improve RVI safety, performance and reliability. It is shown that a perspective of 60 years of operation of RVI components is supported by large Research-Development efforts combined with field experience. (authors)

Pressure vessel rupture within a chamber: the pressure history on the chamber wall

International Nuclear Information System (INIS)

Baum, M.R.

1989-04-01

Generally there is a large number of pressure vessels containing high pressure gas on power stations and chemical plant. In many instances, particularly on power plant, these vessels are within the main building. If a pressure vessel were to fail, the surrounding structures would be exposed to blast loads and the forces resulting from jets of fluid issuing from the breached vessel. In the case where the vessel is in a relatively closed chamber there would also be a general overpressurisation of the chamber. At the design stage it is therefore essential to demonstrate that the plant could be safely shut down in the event of a pressure vessel failure, that is, it must be shown that the chamber will not collapse thus putting the building at risk or hazarding equipment essential for a safe shut down. Such an assessment requires the loads applied to the chamber walls, roof, etc. to be known. (author)

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)

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.

33 CFR 164.78 - Navigation under way: Towing vessels.

Science.gov (United States)

2010-07-01

...) Evaluates the danger of each closing visual or radar contact; (5) Knows and applies the variation and... type of correction; (6) Knows the speed and direction of the current, and the set, drift, and tidal... vessels. 164.78 Section 164.78 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND...

Experience in surveillance of the prestress of concrete reactor vessels in Wylfa nuclear power station

International Nuclear Information System (INIS)

Dawson, P.; Paton, A.A.; Walsh, S.R.

1989-01-01

This paper describes experience gained in the in-service surveillance of the prestressing system for the prestressed concrete reactor vessels (PCRVs) at Wylfa nuclear power station. The paper gives details of results for the prestressing system obtained from the statutory in-service inspection program of the PCRVs. The program includes a detailed examination of a selection of prestressing tendon anchorages, anchorage load checks using a lift-off technique on a one percent sample of tendons and corrosion inspection of samples of prestressing strand and determination of their mechanical properties. The results obtained from the above in-service inspections have shown that the prestressing system continues to function within its design limits

Dosimetry and fluence calculations on french PWR vessels comparisons between experiments and calculations

International Nuclear Information System (INIS)

Nimal, J.C.; Bourdet, L.; Guilleret, J.C.; Hedin, F.

1988-01-01

Fluence and damage calculations on PWR pressure vessels and irradiation test specimens are presented for two types of reactor: the franco-belgian (reactor CHOOZ) and the french reactors (CPY program). Comparisons with measurements are given for activation foils and fission detectors; most of them are about irradiation test specimen locations; comparisons are made for the Chooz plant on vessel stainless steel samplings and in the reactor pit

Margination of Stiffened Red Blood Cells Regulated By Vessel Geometry.

Science.gov (United States)

Chen, Yuanyuan; Li, Donghai; Li, Yongjian; Wan, Jiandi; Li, Jiang; Chen, Haosheng

2017-11-10

Margination of stiffened red blood cells has been implicated in many vascular diseases. Here, we report the margination of stiffened RBCs in vivo, and reveal the crucial role of the vessel geometry in the margination by calculations when the blood is seen as viscoelastic fluid. The vessel-geometry-regulated margination is then confirmed by in vitro experiments in microfluidic devices, and it establishes new insights to cell sorting technology and artificial blood vessel fabrication.

Nonlinear analysis of end slabs in prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Abdulrahman, H.O.

1978-01-01

A procedure for the nonlinear analysis of end slabs is prestressed concrete reactor vessels (PCRVs), based on the finite element method, is presented. The applicability of the procedure to the ultimate load analysis of small-scale models of the primary containment of nuclear reactors is shown. Material nonlinearity only is considered. The procedure utilizes the four-node linear quadrilateral isoparametric element with the choice of incorporating the nonconforming modes. This element is used for modeling the vessel as an axisymmetric solid. Concrete is assumed to be an isotropic material in the elastic range. The compressive stresses are judged according to a special form of the Mohr-Coulomb criterion. The nonlinear problem was solved using a generalized Newton-Raphson procedure. A detailed example problem of a pressure vessel with penetrations is presented. This is followed by a summary of the other cases studied. The solutions obtained match very closely the measured response of the test vessels under increasing internal pressure up to failure. The procedure is thus adequate for the assessment of the ultimate load behavior and failure of actual pressure vessels with a moderate demand on human and computational resources

Trajectory planning of tokamak flexible in-vessel inspection robot

Energy Technology Data Exchange (ETDEWEB)

Wang, Hesheng [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China); Chen, Weidong, E-mail: wdchen@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China); Lai, Yinping; He, Tao [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China)

2015-10-15

Highlights: • A tokamak flexible in-vessel inspection robot is designed. • Two trajectory planning methods are used to ensure the full coverage of the first wall scanning. • The method is tested on a simulated platform of EAST with the flexible in-vessel inspection robot. • Experimental results show the effectiveness of the proposed algorithm. - Abstract: Tokamak flexible in-vessel inspection robot is mainly designed to carry a camera for close observation of the first wall of the vacuum vessel, which is essential for the maintenance of the future tokamak reactor without breaking the working condition of the vacuum vessel. A tokamak flexible in-vessel inspection robot is designed. In order to improve efficiency of the remote maintenance, it is necessary to design a corresponding trajectory planning algorithm to complete the automatic full coverage scanning of the complex tokamak cavity. Two different trajectory planning methods, RS (rough scanning) and FS (fine scanning), according to different demands of the task, are used to ensure the full coverage of the first wall scanning. To quickly locate the damage position, the first trajectory planning method is targeted for quick and wide-ranging scan of the tokamak D-shaped section, and the second one is for careful observation. Furthermore, both of the two different trajectory planning methods can ensure the full coverage of the first wall scanning with an optimal end posture. The method is tested on a simulated platform of EAST (Experimental Advanced Superconducting Tokamak) with the flexible in-vessel inspection robot, and the results show the effectiveness of the proposed algorithm.

Trajectory planning of tokamak flexible in-vessel inspection robot

International Nuclear Information System (INIS)

Wang, Hesheng; Chen, Weidong; Lai, Yinping; He, Tao

2015-01-01

Highlights: • A tokamak flexible in-vessel inspection robot is designed. • Two trajectory planning methods are used to ensure the full coverage of the first wall scanning. • The method is tested on a simulated platform of EAST with the flexible in-vessel inspection robot. • Experimental results show the effectiveness of the proposed algorithm. - Abstract: Tokamak flexible in-vessel inspection robot is mainly designed to carry a camera for close observation of the first wall of the vacuum vessel, which is essential for the maintenance of the future tokamak reactor without breaking the working condition of the vacuum vessel. A tokamak flexible in-vessel inspection robot is designed. In order to improve efficiency of the remote maintenance, it is necessary to design a corresponding trajectory planning algorithm to complete the automatic full coverage scanning of the complex tokamak cavity. Two different trajectory planning methods, RS (rough scanning) and FS (fine scanning), according to different demands of the task, are used to ensure the full coverage of the first wall scanning. To quickly locate the damage position, the first trajectory planning method is targeted for quick and wide-ranging scan of the tokamak D-shaped section, and the second one is for careful observation. Furthermore, both of the two different trajectory planning methods can ensure the full coverage of the first wall scanning with an optimal end posture. The method is tested on a simulated platform of EAST (Experimental Advanced Superconducting Tokamak) with the flexible in-vessel inspection robot, and the results show the effectiveness of the proposed algorithm.

NET in-vessel vehicle system

International Nuclear Information System (INIS)

Jones, H.

1991-02-01

The CFFTP/Spar In-vessel Vehicle System concept for in-vessel remote maintenance of the NET/ITER machine is described. It comprises a curved deployable boom, a vehicle which can travel on the boom and an end effector or work unit mounted on the vehicle. The stowed boom, vehicle, and work unit are inserted via the equatorial access port of the torus. Following insertion the boom is deployed and locked in place. The vehicle may then travel along the boom to transport the work unit to any desired location. A novel feature of the concept is the deployable boom. When fully deployed, it closely resembles a conventional curved truss structure in configuration and characteristics. However, the joints of the truss structure are hinged so that it can fold into a compact package, of less than 20% of deployed volume for storage, transportation and insertion into the torus. A full-scale 2-metre long section of this boom was produced for demonstration purposes. As part of the concept definition the work unit for divertor handling was studied to demonstrate that large payloads could be manipulated within the confines of the torus using the in-vessel vehicle system. Principal advantages of the IVVS are its high load capacity and rigidity, low weight and stowed volume, simplicity of control and operation, and its relatively high speed of transportation

User`s guide for the irradiation of experiments in the FTR. Revision 3

Energy Technology Data Exchange (ETDEWEB)

NONE

1990-07-01

This document provides Revision 3 updates the FTR Users Guide. Revision 3 updates Appendix 1 (FFTF Technical Specifications) to include the following: Documentation of the acceptability of handling metal fuel assemblies in the Closed Loop Ex-Vessel Handling Machine (CLEM) and storing them in the Interim Decay Storage (IDS) vessel. Reactivity limit version (utilizing existing FSAR analysis bounds) to allow for the larger beta-effective associated with the addition of enriched uranium metal and oxide experiments to the core. Operational temperature limits for Open Test Assemblies (OTAs) have been expanded to differentiate between 40-foot experiment test articles, 28-foot Post Irradiation Open Test Assemblies (PIOTAs) and the 28-foot Loose Parts Monitor Assemblies (LMPAs) operating under FFTF core Engineering cognizance.

Simulation of Diffusive Lithium Evaporation Onto the NSTX Vessel Walls

International Nuclear Information System (INIS)

Stotler, D.P.; Skinner, C.H.; Blanchard, W.R.; Krstic, P.S.; Kugel, H.W.; Schneider, H.; Zakharov, L.E.

2010-01-01

A model for simulating the diffusive evaporation of lithium into a helium filled NSTX vacuum vessel is described and validated against an initial set of deposition experiments. The DEGAS 2 based model consists of a three-dimensional representation of the vacuum vessel, the elastic scattering process, and a kinetic description of the evaporated atoms. Additional assumptions are required to account for deuterium out-gassing during the validation experiments. The model agrees with the data over a range of pressures to within the estimated uncertainties. Suggestions are made for more discriminating experiments that will lead to an improved model.

The DOS 1 neutron dosimetry experiment at the HB-4-A key 7 surveillance site on the HFIR pressure vessel

International Nuclear Information System (INIS)

Farrell, K.; Kam, F.B.; Baldwin, C.A.

1994-01-01

A comprehensive neutron dosimetry experiment was made at one of the prime surveillance sites at the High Flux Isotope Reactor (HFIR) pressure vessel to aid radiation embrittlement studies of the vessel and to benchmark neutron transport calculations. The thermal neutron flux at the key 7, position 5 site was found, from measurements of radioactivation of four cobalt wires and four silver wires, to be 2.4 x 10 12 n·m -2 ·s -1 . The thermal flux derived from two helium accumulation monitors was 2.3 x 10 12 n·m -2 · -1 . The thermal flux estimated by neutron transport calculations was 3.7 x 10 12 n·m -2 s -1 . The fast flux, >1 MeV, determined from two nickel activation wires, was 1.5 x 10 12 n·m -2 ·s -1 , in keeping with values obtained earlier from stainless steel surveillance monitors and with a computed value of 1.2 x 10 13 n·m -2 · -1 . The fast fluxes given by two reaction-product-type monitors, neptunium-237 and beryllium, were 2.6 x 10 13 n·m -2 ·s -1 and 2.2 x 10 13 n·m -2 s -1 , respectively. Follow-up experiments indicate that these latter high values of fast flux are reproducible but are false; they are due to the creation of greater levels of reaction products by photonuclear events induced by an exceptionally high ratio of gamma flux to fast neutron flux at the vessel

Small-target leak detection for a closed vessel via infrared image sequences

Science.gov (United States)

Zhao, Ling; Yang, Hongjiu

2017-03-01

This paper focus on a leak diagnosis and localization method based on infrared image sequences. Some problems on high probability of false warning and negative affect for marginal information are solved by leak detection. An experimental model is established for leak diagnosis and localization on infrared image sequences. The differential background prediction is presented to eliminate the negative affect of marginal information on test vessel based on a kernel regression method. A pipeline filter based on layering voting is designed to reduce probability of leak point false warning. A synthesize leak diagnosis and localization algorithm is proposed based on infrared image sequences. The effectiveness and potential are shown for developed techniques through experimental results.

Vessel generator noise as a settlement cue for marine biofouling species.

Science.gov (United States)

McDonald, J I; Wilkens, S L; Stanley, J A; Jeffs, A G

2014-01-01

Underwater noise is increasing globally, largely due to increased vessel numbers and international ocean trade. Vessels are also a major vector for translocation of non-indigenous marine species which can have serious implications for biosecurity. The possibility that underwater noise from fishing vessels may promote settlement of biofouling on hulls was investigated for the ascidian Ciona intestinalis. Spatial differences in biofouling appear to be correlated with spatial differences in the intensity and frequency of the noise emitted by the vessel's generator. This correlation was confirmed in laboratory experiments where C. intestinalis larvae showed significantly faster settlement and metamorphosis when exposed to the underwater noise produced by the vessel generator. Larval survival rates were also significantly higher in treatments exposed to vessel generator noise. Enhanced settlement attributable to vessel generator noise may indicate that vessels not only provide a suitable fouling substratum, but vessels running generators may be attracting larvae and enhancing their survival and growth.

In vessel retention for VVER 1000 - Experimental work

International Nuclear Information System (INIS)

Batek, D.

2015-01-01

After Fukushima accident, the nuclear community realized that it is necessary to have strategy and solution for severe accident management. In Vessel Retention (IVR) of corium is an important strategy to mitigate the consequences of a severe accident. In this poster the author reviews the present status of experimental works made by UJV (Czech Republic) from 2012 until now, on the IVR strategy specifically applied for the VVER 1000 unit. The BESTH 1 experiment was prepared to test the behavior of the RPV (Reactor Pressure Vessel) surface under 2 configurations: clean and corroded. BESTH 2 experiment is a modification of BESTH 1 experiment in order to get greater thermal fluxes. The BESTH 3 facility is a large scale experiment that is under extensive design (2016-2017) whose main objective will be to investigate the results of vast analytical works made by experts with specialization of severe accident phenomenology

Study on the combustion behavior of radiolytically generated hydrogen explosion in small scale annular vessels at the reprocessing plant

International Nuclear Information System (INIS)

Kudo, Tatsuya; Tamauchi, Yoshikazu; Arai, Nobuyuki; Dai, Wenbin; Sakaihara, Motohiro; Kanehira, Osamu

2017-01-01

Hydrogen is generated by radiolysis of water, etc. in process vessels in reprocessing plant. Usually, the hydrogen is scavenged by compressed air into vessels to prevent hydrogen explosion. When an earthquake beyond design based occurs, for example, the compressed air may stop and the hydrogen starts accumulating in the vessels, and under this condition, an ignition source might set off hydrogen explosion. Therefore, the explosion derived by the radiolytically generated hydrogen is designated as one of severe accidents on Rokkasho Reprocessing Plant in new regulatory requirements. It is important to understand the combustion behavior of hydrogen explosion inside a vessel for consideration of safety measures against the severe accident, because the influences of detonation are not considered in the design basis of vessels. Especially, the investigations about the combustion behavior which considered influence of interior obstacles inside the vessel are not performed yet. In order to investigate the combustion behavior comprehensively, explosion experiment, combustion analysis and structural analysis are carried out using the representative vessels (small scale annular vessel, small scale plate vessel, large scale annular vessel and large scale cylindrical vessel) selected from Rokkasho Reprocessing Plant. In this paper, the results of experiments and analysis of small scale annular vessel (as one of representative vessel, imitated a pulsed column in the reprocessing plant) are reported. As imitated vessels, three vessels are manufactured with different interior obstacle arrangements as follows, A) cylindrical obstacles are faithfully reproduced and are arranged based on the actual vessel, B) cylindrical obstacles are arranged more densely than the actual vessel, and C) there are no obstacles inside the vessel. Experiments of hydrogen explosion are performed under condition of stoichiometric hydrogen-air ratio (premixed hydrogen-air is used). As a result of

Experience of in-service surveillance and monitoring of prestressed concrete pressure vessels for nuclear reactors

International Nuclear Information System (INIS)

Irving, J.; Smith, J.R.; Eadie, D.McD.; Hornby, I.W.

1976-01-01

Details are given of the statutory requirements for the inspection of prestressed concrete pressure vessels in the United Kingdom, with particular emphasis on the prestressing system. The results of periodic examinations under the Licencing Conditions of the Oldbury and Wylfa vessels are presented and discussed in relation to design expectations and future requirements. Strain, moisture and temperature records obtained from the Oldbury PCPV's over a 10 year period are compared with prediction and new developments in vessel instrumentation are discussed. (author)

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.

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.

Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

International Nuclear Information System (INIS)

Cheverton, R.D.

1976-09-01

The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall

PLANNING VESSEL BODY SECTION PRODUCTION

Directory of Open Access Journals (Sweden)

A. G. Grivachevsky

2015-01-01

Full Text Available A problem of planning production of a vessel body section is considered. The problem is reduced to the classic Johnson’s tree-machine flow-shop scheduling problem. A genetic algorithm and computer experiment to compare efficiency of this algorithm and the algorithm of full enumeration are described.

PWR reactor vessel in-service-inspection according to RSEM

International Nuclear Information System (INIS)

Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul

2006-01-01

Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high requirements of

Equation of state of an ideal gas with nonergodic behavior in two connected vessels.

Science.gov (United States)

Naplekov, D M; Semynozhenko, V P; Yanovsky, V V

2014-01-01

We consider a two-dimensional collisionless ideal gas in the two vessels connected through a small hole. One of them is a well-behaved chaotic billiard, another one is known to be nonergodic. A significant part of the second vessel's phase space is occupied by an island of stability. In the works of Zaslavsky and coauthors, distribution of Poincaré recurrence times in similar systems was considered. We study the gas pressure in the vessels; it is uniform in the first vessel and not uniform in second one. An equation of the gas state in the first vessel is obtained. Despite the very different phase-space structure, behavior of the second vessel is found to be very close to the behavior of a good ergodic billiard but of different volume. The equation of state differs from the ordinary equation of ideal gas state by an amendment to the vessel's volume. Correlation of this amendment with a share of the phase space under remaining intact islands of stability is shown.

Conceptual design of EAST flexible in-vessel inspection system

International Nuclear Information System (INIS)

Peng, X.B.; Song, Y.T.; Li, C.C.; Lei, M.Z.; Li, G.

2010-01-01

Remote handling technology, especially the flexible in-vessel inspection system (FIVIS) without breaking the working condition of the vacuum vessel, has been identified as one major challenge on the maintenance for the future tokamak fusion reactor. The FIVIS introduced here is specially developed for EAST superconducting tokamak that has actively cooled plasma facing components (PFCs). It aims flexible close-up inspection of EAST PFCs to help the understanding of operation issues that could occur in the vacuum vessel. This paper resumes the preliminary work of the FIVIS project, including the requirement analysis and the development of the conceptual design. The FIVIS consists out of a long reach multi-articulated manipulator and a process tool. The manipulator has a modular design for its subsystems and can reach all areas of the first wall in the distance of 15 mm and in the range of ±90 o along toroidal direction. It will be folded and hidden in the designated horizontal port during plasma discharge period.

Physico-Chemistry and Corium Properties for In-Vessel Retention

International Nuclear Information System (INIS)

Froment, K.; Seiler, J.M.; Gueneau, C.; Dauvois, V.; Barbier, F.; Bellon, M.; Tourasse, M.; Ducros, G.; Cognet, G.; Sudreau, F.

1999-01-01

This paper focuses on some important aspects of consequences of material behaviour and interactions on in-vessel retention capabilities. It discusses the behaviour of corium oxide mixtures at elevated temperatures (miscibility gap and density effects, separation due to density effects in the solid-liquid mixture according to the analysis of the Rasplav experiment results), and then the interaction between metallic layer and vessel wall (physical-chemical interaction of corium with the carbon steel vessel wall, migration of low melting point metallic elements in the solid vessel wall). It proposes a mode for the calculation of melt viscosity (liquid phase viscosity and viscosity in the solidification range), addresses the issue of barium release and residual power and of distribution of the residual power in an oxidic corium

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

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

EDF experience with {open_quotes}hot spot{close_quotes} management

Energy Technology Data Exchange (ETDEWEB)

Guio, J.M. de [Blayais Nuclear Power Plant, St. Ciers (France)

1995-03-01

During the past few years, {open_quotes}hot spots{close_quotes} due to the presence of particles of metal activated during their migration through the reactor core, have been detected at several French pressurized water reactor (PWR) units. These {open_quotes}hot spots,{close_quotes} which generate very high dose rates (from about 10 Gy/h to 200 G/h) are a significant factor in increase occupational exposures during outrates. Of particular concern are the difficult cases which prolong outage duration and increase the volume of radiological waste. Confronted with this situation, Electricite de France (EDF) has set up a national research group, as part of its ALARA program, to establish procedures and techniques to avoid, detect, and eliminate of hot spots. In particular, specific processes have been developed to eliminate these hot spots which are most costly in terms of occupational exposure due to the need for reactor maintenance. This paper sets out the general approach adopted at EDF so far to cope with the problem of hot spots, illustrated by experience at Blayais 3 and 4.

Neutron fluence determination for light water reactor pressure vessels

International Nuclear Information System (INIS)

Gold, R.

1994-01-01

A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

Multi-level deep supervised networks for retinal vessel segmentation.

Science.gov (United States)

Mo, Juan; Zhang, Lei

2017-12-01

Changes in the appearance of retinal blood vessels are an important indicator for various ophthalmologic and cardiovascular diseases, including diabetes, hypertension, arteriosclerosis, and choroidal neovascularization. Vessel segmentation from retinal images is very challenging because of low blood vessel contrast, intricate vessel topology, and the presence of pathologies such as microaneurysms and hemorrhages. To overcome these challenges, we propose a neural network-based method for vessel segmentation. A deep supervised fully convolutional network is developed by leveraging multi-level hierarchical features of the deep networks. To improve the discriminative capability of features in lower layers of the deep network and guide the gradient back propagation to overcome gradient vanishing, deep supervision with auxiliary classifiers is incorporated in some intermediate layers of the network. Moreover, the transferred knowledge learned from other domains is used to alleviate the issue of insufficient medical training data. The proposed approach does not rely on hand-crafted features and needs no problem-specific preprocessing or postprocessing, which reduces the impact of subjective factors. We evaluate the proposed method on three publicly available databases, the DRIVE, STARE, and CHASE_DB1 databases. Extensive experiments demonstrate that our approach achieves better or comparable performance to state-of-the-art methods with a much faster processing speed, making it suitable for real-world clinical applications. The results of cross-training experiments demonstrate its robustness with respect to the training set. The proposed approach segments retinal vessels accurately with a much faster processing speed and can be easily applied to other biomedical segmentation tasks.

Vascular remodeling: A redox-modulated mechanism of vessel caliber regulation.

Science.gov (United States)

Tanaka, Leonardo Y; Laurindo, Francisco R M

2017-08-01

Vascular remodeling, i.e. whole-vessel structural reshaping, determines lumen caliber in (patho)physiology. Here we review mechanisms underlying vessel remodeling, with emphasis in redox regulation. First, we discuss confusing terminology and focus on strictu sensu remodeling. Second, we propose a mechanobiological remodeling paradigm based on the concept of tensional homeostasis as a setpoint regulator. We first focus on shear-mediated models as prototypes of remodeling closely dominated by highly redox-sensitive endothelial function. More detailed discussions focus on mechanosensors, integrins, extracellular matrix, cytoskeleton and inflammatory pathways as potential of mechanisms potentially coupling tensional homeostasis to redox regulation. Further discussion of remodeling associated with atherosclerosis and injury repair highlights important aspects of redox vascular responses. While neointima formation has not shown consistent responsiveness to antioxidants, vessel remodeling has been more clearly responsive, indicating that despite the multilevel redox signaling pathways, there is a coordinated response of the whole vessel. Among mechanisms that may orchestrate redox pathways, we discuss roles of superoxide dismutase activity and extracellular protein disulfide isomerase. We then discuss redox modulation of aneurysms, a special case of expansive remodeling. We propose that the redox modulation of vascular remodeling may reflect (1) remodeling pathophysiology is dominated by a particularly redox-sensitive cell type, e.g., endothelial cells (2) redox pathways are temporospatially coordinated at an organ level across distinct cellular and acellular structures or (3) the tensional homeostasis setpoint is closely connected to redox signaling. The mechanobiological/redox model discussed here can be a basis for improved understanding of remodeling and helps clarifying mechanisms underlying prevalent hard-to-treat diseases. Copyright © 2017 Elsevier Inc. All

Foundamental characteristics of layered pressure vessel

International Nuclear Information System (INIS)

Moriwaki, Yoshikazu; Fugino, Masayuki; Shimizu, Yasuhiro; Nakamura, Takeshi

1978-01-01

Pressure vessels become larger and the working pressure become higher with the remarkable development of petroleum, chemical, thermal power generation and atomic energy industries. Multi-layered pressure vessels can be manufactured cheaply without large installations, and large wall thickness can be made, therefore they are suitable for large pressure vessels. The stress and deformation behaviors of such vessels are very complex because of the effect of frictional force working between layers. In this study, the phenomena arising in multiple layers and the difference as compared with single wall were studied fundamentally as one step for analyzing multi-layered pressure vessels as a whole. Finite element technique was employed as the analyzing method, and the behavior of multiple layers was analyzed, regarding it as multiple contact problem. The behavior of multiple layers seems to appear conspicuously in case of bending load, therefore the basic characteristics regarding bending were examined. The evaluation of interfacial stiffness was carried out by experiment. The computer program for analyzing multiple contact problem was developed. In order to examine the validity of the program, comparison with the analytical solution heretofore and the result of calculation by finite element technique was carried out. Moreover, the experimental proof with multi-layered models was made. The frictional force between layers hardly contributes to the stiffness. (Kako, I.)

Characterizing experiments of the PPOOLEX test facility

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-15

This report summarizes the results of the characterizing test series in 2007 with the scaled down PPOOLEX facility designed and constructed at Lappeenranta University of Technology. Air and steam/air mixture was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool (wet well). Altogether eight air and four steam/air mixture experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the general behavior of the facility and the performance of basic instrumentation. Proper operation of automation, control and safety systems was also tested. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. The facility is equipped with high frequency measurements for capturing different aspects of the investigated phenomena. The general behavior of the PPOOLEX facility differs significantly from that of the previous POOLEX facility because of the closed two-compartment structure of the test vessel. Heat-up by several tens of degrees due to compression in both compartments was the most obvious evidence of this. Temperatures also stratified. Condensation oscillations and chugging phenomenon were encountered in those tests where the fraction of non-condensables had time to decrease significantly. A radical change from smooth condensation behavior to oscillating one occurred quite abruptly when the air fraction of the blowdown pipe flow dropped close to zero. The experiments again demonstrated the strong diminishing effect that noncondensable gases have on dynamic unsteady loadings experienced by submerged pool structures. BWR containment like behavior related to the beginning of a postulated steam line break accident was observed in the PPOOLEX test facility during the steam/air mixture experiments. The most important task of the research project, to produce experimental data for code simulation purposes, can be

Slideline verification for multilayer pressure vessel and piping analysis including tangential motion

International Nuclear Information System (INIS)

Van Gulick, L.A.

1984-01-01

Nonlinear finite element method (FEM) computer codes with slideline algorithm implementations should be useful for the analysis of prestressed multilayer pressure vessels and piping. This paper presents closed form solutions including the effects of tangential motion useful for verifying slideline implementations for this purpose. The solutions describe stresses and displacements of a long internally pressurized elastic-plastic cylinder initially separated from an elastic outer cylinder by a uniform gap. Comparison of closed form and FEM results evaluates the usefulness of the closed form solution and the validity of the sideline implementation used

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

International Nuclear Information System (INIS)

Seiger, H.; Engl, G.

1976-01-01

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

Stress criteria for nuclear vessel concrete

International Nuclear Information System (INIS)

Costes, D.

1975-01-01

Concrete nuclear vessels are submitted to prestressing forces which limit tensile stresses in concrete when the vessel is under pressure with thermal gradients. Hence, the most severe conditions for concrete appear when the vessel is prestressed and not submitted to internal pressure. The triaxial states of stress in the concrete may be computed postulating elastic or other behavior and compared with safe limits obtained from rupture tests and fatigue tests. The first part of the paper, recalls experimental rupture results and the acceptability procedures currently used. Criteria founded on the lemniscoid surfaces are proposed, parameters for which are obtained by various tests and safety considerations. In the second part, rupture tests are reported on small, thick, cylindrical vessels submitted to external hydraulic pressure simulating prestressing forces. Materials used are plain concrete, microconcrete, marble and graphite. The strengths obtained are much higher than those which could be elastically computed, triaxial rupture states being provided by previous experiments. Such results may be due to a plastic stress redistribution before fracture and to stabilizing effects of stress gradients around the more stressed areas. Fatigue tests by external hydraulic loading are reported [fr

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

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.

Interaction between U(VI) and Fe(II) in aqueous solution under anaerobic conditions. Closed system experiments

International Nuclear Information System (INIS)

Myllykylae, E.; Ollila, K.

2011-01-01

The aim of these experiments is to investigate the potential reduction of U(VI) carbonate and hydroxide complexes by aqueous Fe(II). This reduction phenomenon could be important under the disposal conditions of spent fuel. If groundwater enters the copper/iron canister, alpha radiolysis of the water may locally induce oxidizing conditions on the surface of UO 2 fuel, leading to the dissolution of UO 2 as more soluble U(VI) species. A potential reducing agent in the intruding water is Fe(II)(aq) from anaerobic corrosion of the copper/iron canister. The reduction of U(VI) to U(IV) would substantially decrease the solubility of U as well as co-precipitate other actinides and radionuclides. The interaction experiments were conducted in 0.01 M NaCl and 0.002 M NaHCO 3 solutions using an initial uranium concentration of either 8.4 x 10 -8 or 4.2 x 10 -7 mol/L with an initial Fe(II) concentration of 1.8 x 10 -6 in the NaCl solutions and 1.3 x 10 -6 mol/L in the NaHCO 3 solutions. Only after an equilibration period for U(VI) complexation was Fe(II) added to the solutions. The reaction times varied from 1 week to 5 months. For extra protection against O 2 , even inside a glove-box (N 2 atmosphere), the plastic reaction vessels were closed in metallic containers. The concentrations of U, Fe TOT and Fe(II) were analysed as a function of time for unfiltered, micro- and ultrafiltered samples. In addition, the precipitate on the ultrafilters was analysed with ESEM-EDS. The evolution of pH and Eh values was measured. The oxidation state of U in solution was preliminarily analysed for chosen periods. The results of the tests in 0.01 M NaCl showed an initial rapid decrease in U concentration after the addition of Fe(II) to the solution. The U found on test vessel walls at the end of the reaction periods, as well as the ESEM-EDS analyses of the filtered precipitates from the test solutions, showed that precipitation of U had occurred. The oxidation state analyses showed the presence

Design description of the vacuum vessel for the Advanced Toroidal Facility

International Nuclear Information System (INIS)

Chipley, K.K.; Nelson, B.E.; Vinyard, L.M.; Williamson, D.F.

1983-01-01

The Advanced Toroidal Facility (ATF) will be a stellarator experiment to investigate improvements in toroidal confinement. The vacuum vessel for this facility will provide the appropriate evacuated region for plasma containment within the helical field (HF) coils. The vessel is designed to provide the maximum reasonable volume inside the HF coils and to provide the maximum reasonable access for future diagnostics. The vacuum vessel design is at an early phase and all of the details have not been completed. The heat transfer analysis and stress analysis completed during the conceptual design indicate that the vessel will not change drastically

Performance of core exit thermocouple for PWR accident management action in vessel top break LOCA simulation experiment at OECD/NEA ROSA project

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

2009-01-01

Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary side in case of core temperature excursion. Test 6-1 is first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reason of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Sarkimo, M; Rintamac, R

1988-12-31

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

Dynamics of bubble collapse under vessel confinement in 2D hydrodynamic experiments

Science.gov (United States)

Shpuntova, Galina; Austin, Joanna

2013-11-01

One trauma mechanism in biomedical treatment techniques based on the application of cumulative pressure pulses generated either externally (as in shock-wave lithotripsy) or internally (by laser-induced plasma) is the collapse of voids. However, prediction of void-collapse driven tissue damage is a challenging problem, involving complex and dynamic thermomechanical processes in a heterogeneous material. We carry out a series of model experiments to investigate the hydrodynamic processes of voids collapsing under dynamic loading in configurations designed to model cavitation with vessel confinement. The baseline case of void collapse near a single interface is also examined. Thin sheets of tissue-surrogate polymer materials with varying acoustic impedance are used to create one or two parallel material interfaces near the void. Shadowgraph photography and two-color, single-frame particle image velocimetry quantify bubble collapse dynamics including jetting, interface dynamics and penetration, and the response of the surrounding material. Research supported by NSF Award #0954769, ``CAREER: Dynamics and damage of void collapse in biological materials under stress wave loading.''

A powerful methodology for reactor vessel pressurized thermal shock analysis

International Nuclear Information System (INIS)

Boucau, J.; Mager, T.

1994-01-01

The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). More specifically, the review of the old WWER-type of reactors (WWER 440/230) has indicated a sensitive behaviour to neutron embrittlement. This led already to some remedial actions including safety injection water preheating or vessel annealing. Such measures are usually taken based on the analysis of a selected number of conservative PTS events. Consideration of all postulated cooldown events would draw attention to the impact of operator action and control system effects on reactor vessel PTS. Westinghouse has developed a methodology which couples event sequence analysis with probabilistic fracture mechanics analyses, to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. Once the event sequences of concern are identified, detailed deterministic thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. The results of these analyses can then be used to better define further modifications in vessel and plant system design and to operating procedures. The purpose of the present paper will be to describe this methodology and to show its benefits for decision making. (author). 1 ref., 3 figs

Experimental studies of oxidic molten corium-vessel steel interaction

International Nuclear Information System (INIS)

Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Lopukh, D.B.; Petrov, Yu.B.; Petchenkov, A.Yu.; Kulagin, I.V.; Granovsky, V.S.; Kovtunova, S.V.; Martinov, V.V.; Gusarov, V.V.

2001-01-01

The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere

Experimental studies of oxidic molten corium-vessel steel interaction

Energy Technology Data Exchange (ETDEWEB)

Bechta, S.V. E-mail: niti-npc@sbor.net; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Lopukh, D.B.; Petrov, Yu.B.; Petchenkov, A.Yu.; Kulagin, I.V.; Granovsky, V.S.; Kovtunova, S.V.; Martinov, V.V.; Gusarov, V.V

2001-12-01

The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere.

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

Irradiation embrittlement of pressure vessel steels

International Nuclear Information System (INIS)

Brumovsky, M.; Vacek, M.

1975-01-01

A Standard Research Programme on Irradiation Embrittlement of Pressure Vessel Steels was approved by the Coordinating Meeting on the 12th May 1972 at the Working Group on Engineering Aspects of Irradiation Embrittlement of Pressure Vessel Steels. This Working Group was set up by the International Atomic Energy Agency in Vienna. Seven countries with their research institutes agreed on doing irradiation experiments according to the approved programme on steel A533 B from the U.S. HSST Programme. The Czechoslovak contribution covering tensile and impact testing of non-irradiated steel and steel irradiated at 280degC to 1.3 x 10 23 n/m 2 (E above 1 MeV) is presented in this report. As an additional part the same set of experiments was carried out on two additional steels - A 542 and A 543, made in SKODA Works for comparison of their irradiation embrittlement and hardening with A533 B steel. (author)

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...

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...

DHCVIM - a direct heating containment vessel interactions module: applications to Sandia National Laboratories Surtsey experiments

International Nuclear Information System (INIS)

Ginsberg, T.; Tutu, N.K.

1987-01-01

Direct containment heating is the mechanism of severe nuclear reactor accident containment loading that results from transfer of thermal and chemical energy from high-temperature, finely divided, molten core material to the containment atmosphere. The direct heating containment vessel interactions module (DHCVIM) has been developed at Brookhaven National Laboratory to model the mechanisms of containment loading resulting from the direct heating accident sequence. The calculational procedure is being used at present to model the Sandia National Laboratories one-tenth-scale Surtsey direct containment heating experiments. The objective of the code is to provide a test bed for detailed modeling of various aspects of the thermal, chemical, and hydrodynamic interactions that are expected to occur in three regions of a containment building: reactor cavity, intermediate subcompartments, and containment dome. Major emphasis is placed on the description of reactor cavity dynamics. This paper summarizes the modeling principles that are incorporated in DHCVIM and presents a prediction of the Surtsey Test DCH-2 that was made prior to execution of the experiment

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

Vessel Enhancement and Segmentation of 4D CT Lung Image Using Stick Tensor Voting

Science.gov (United States)

Cong, Tan; Hao, Yang; Jingli, Shi; Xuan, Yang

2016-12-01

Vessel enhancement and segmentation plays a significant role in medical image analysis. This paper proposes a novel vessel enhancement and segmentation method for 4D CT lung image using stick tensor voting algorithm, which focuses on addressing the vessel distortion issue of vessel enhancement diffusion (VED) method. Furthermore, the enhanced results are easily segmented using level-set segmentation. In our method, firstly, vessels are filtered using Frangi's filter to reduce intrapulmonary noises and extract rough blood vessels. Secondly, stick tensor voting algorithm is employed to estimate the correct direction along the vessel. Then the estimated direction along the vessel is used as the anisotropic diffusion direction of vessel in VED algorithm, which makes the intensity diffusion of points locating at the vessel wall be consistent with the directions of vessels and enhance the tubular features of vessels. Finally, vessels can be extracted from the enhanced image by applying level-set segmentation method. A number of experiments results show that our method outperforms traditional VED method in vessel enhancement and results in satisfied segmented vessels.

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

Corrosion of vessel steel during its interaction with molten corium

International Nuclear Information System (INIS)

Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Lopukh, D.B.; Gusarov, V.V.; Martinov, A.P.; Martinov, V.V.; Fieg, G.; Tromm, W.; Bottomley, D.; Tuomisto, H.

2006-01-01

An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments have been conducted on 'Rasplav-2' test facility and followed up with physico-chemical and metallographic analyses of melt samples and corium-specimen ingots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere above the melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate or corrosion depth of vessel steel in conditions simulated by the experiments

Corrosion of vessel steel during its interaction with molten corium

Energy Technology Data Exchange (ETDEWEB)

Bechta, S.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation)]. E-mail: bechta@sbor.spb.su; Khabensky, V.B. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Vitol, S.A. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Krushinov, E.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Granovsky, V.S. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Lopukh, D.B. [SPb Electrotechnical University (SpbGETU), Professor Popov str., b.5/3, 197376 St. Petersburg (Russian Federation); Gusarov, V.V. [Institute of Silicate Chemistry of Russian Academy of Science (ISC of RAS), Odoevsky str., b. 24/2, 199155 St. Petersburg (Russian Federation); Martinov, A.P. [SPb Electrotechnical University (SpbGETU), Professor Popov str., b.5/3, 197376 St. Petersburg (Russian Federation); Martinov, V.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Fieg, G. [Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik, Postfach 3640, D-78021 Karlsruhe (Germany); Tromm, W. [Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik, Postfach 3640, D-78021 Karlsruhe (Germany); Bottomley, D. [Europaeische Kommission, General Direktion GFS, Institut fuer Transurane (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Tuomisto, H. [Fortum Engineering Ltd. 00048 FORTUM, Rajatorpantie 8, Vantaa (Finland)

2006-07-15

An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments have been conducted on 'Rasplav-2' test facility and followed up with physico-chemical and metallographic analyses of melt samples and corium-specimen ingots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere above the melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate or corrosion depth of vessel steel in conditions simulated by the experiments.

Structural analysis of the JT-60SA cryostat vessel body

Energy Technology Data Exchange (ETDEWEB)

Botija, José, E-mail: jose.botija@ciemat.es [Association EURATOM – CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Alonso, Javier; Fernández, Pilar; Medrano, Mercedes; Ramos, Francisco; Rincon, Esther; Soleto, Alfonso [Association EURATOM – CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Davis, Sam; Di Pietro, Enrico; Tomarchio, Valerio [Fusion for Energy, JT-60SA European Home Team, 85748 Garching bei Munchen (Germany); Masaki, Kei; Sakasai, Akira; Shibama, Yusuke [JAEA – Japan Atomic Energy Agency, Naka Fusion Institute, Ibaraki 311-0193 (Japan)

2013-10-15

Highlights: ► Structural analysis to validate the JT-60SA cryostat vessel body design. ► Design code ASME 2007 “Boiler and Pressure Vessel Code. Section VIII”. ► First buckling mode: load multiplier of 10.644, higher than the minimum factor 4.7. ► Elastic and elastic–plastic stress analysis meets ASME against plastic collapse. ► Bolted fasteners have been analyzed showing small gaps closed by strong welding. -- Abstract: The JT-60SA cryostat is a stainless steel vacuum vessel (14 m diameter, 16 m height) which encloses the Tokamak providing the vacuum environment (10{sup −3} Pa) necessary to limit the transmission of thermal loads to the components at cryogenic temperature. It must withstand both external atmospheric pressure during normal operation and internal overpressure in case of an accident. The paper summarizes the structural analyses performed in order to validate the JT-60SA cryostat vessel body design. It comprises several analyses: a buckling analysis to demonstrate stability under the external pressure; an elastic and an elastic–plastic stress analysis according to ASME VIII rules, to evaluate resistance to plastic collapse including localized stress concentrations; and, finally, a detailed analysis with bolted fasteners in order to evaluate the behavior of the flanges, assuring the integrity of the vacuum sealing welds of the cryostat vessel body.

Application of radioisotope tracer techniques in evaluation of irradiation vessel of flue gas treatment system

International Nuclear Information System (INIS)

Joon-Ha Jin; Myun-Joo Lee; Sung-Hee Jung; Young-Chang Nho

1998-01-01

The proper design of the irradiation vessel of electron beam flue gases treatment plant and resultant optimum gas flow pattern is a very important factor to get a high removal efficiency of toxic materials from flue gases. Radioisotope tracer experiments were conducted to study the residence time distribution of gas flow in a cylindrical irradiation vessel. A few mCi of gaseous radioisotope tracer Ar-41 was injected to the upstream of the vessel and the input and output response were measured with two NaI scintillation detectors. The same experiment was conducted after the modification of the vessel by introducing 4 baffles. The experimental data were analyzed to calculate mean residence times and mixing characteristics of each system using the residence time distribution (RTD) analysis software. A method to estimate pollutant removal efficiencies of an irradiation vessel from the residence time distributions measured by radiotracer experiments was suggested. The analytical results were compared to evaluate the effect of the baffles on the removal efficiency of the plant

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

International Nuclear Information System (INIS)

Mitchell, D.H.

1988-06-01

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

Experimental study of laminar and turbulent flame speed of a spherical flame in a fan-stirred closed vessel for hydrogen safety application

Energy Technology Data Exchange (ETDEWEB)

Goulier, J. [Institut de Combustion, Aérothermique, Réactivité et Environnement, CNRS-ICARE (France); Institut de Radioprotection et de Sûreté Nucléaire (IRSN) (France); Chaumeix, N., E-mail: chaumeix@cnrs-orleans.fr [Institut de Combustion, Aérothermique, Réactivité et Environnement, CNRS-ICARE (France); Halter, F. [Institut de Combustion, Aérothermique, Réactivité et Environnement, CNRS-ICARE (France); Meynet, N.; Bentaïb, A. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN) (France)

2017-02-15

The aim of this paper is to report new experimental results on the effect of turbulence on the propagation speed of hydrogen/air flames. To do so, a new experimental setup, called the spherical bomb, has been designed and built at CNRS-ICARE laboratory. With this new setup, the effect of a given and well-characterized turbulence intensity on the increase of hydrogen/air flame speed can be investigated. This new facility consists of a spherical vessel equipped (563 mm internal diameter) equipped with 8 motors which are linked to fans inside the bomb. Fan actuation induces the generation of a turbulent flow inside the vessel prior to any ignition. The spherical bomb is equipped with 4 quartz windows (200 mm optical diameter) that allow the use of a Particle Image Velocimetry diagnostic in order to characterize the turbulence level inside the bomb. The flame propagation was recorded using a high speed camera at 19,002 frames per second. These experiments were performed for lean to stoichiometric hydrogen/air mixtures (16–20% of H{sub 2} in air), initially at ambient temperature and pressure, and for a rotation speed from 1000 to 5000 rpm. The PIV measurements showed that a homogeneous and isotropic turbulence is created with a fluctuation speed that can reach 4 m/s at 5000 rpm.

Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

Energy Technology Data Exchange (ETDEWEB)

Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

2013-07-01

Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle

Optimized design of an ex-vessel cooling thermosyphon for decay heat removal in SFR

International Nuclear Information System (INIS)

Choi, Jae Young; Jeong, Yong Hoon; Song, Sub Lee; Chang, Soon Heung

2017-01-01

Passive decay heat removal and sodium fire are two major key issues of nuclear safety in sodium-cooled fast reactor (SFR). Several decay heat removal systems (DHR) were suggested for SFR around the world so far. Those DHRS mainly classified into two concepts: Direct reactor cooling system and ex-vessel cooling system. Direct reactor cooling method represented by PDHRS from PGSFR has disadvantages on its additional in-vessel structure and potential sodium fire risk due to the sodium-filled heat exchanger exposed to air. Contrastively, ex-vessel cooling method represented by RVACS from PRISM has low decay heat removal performance, which cannot be applicable to large scale reactors, generally over 1000 MWth. No passive DHRSs which can solve both side of disadvantages has been suggested yet. The goal of this study was to propose ex-vessel cooling system using two-phase closed thermosyphon to compensate the disadvantages of the past DHRSs. Reference reactor was Innovative SFR (iSFR), a pool-type SFR designed by KAIST and featured by extended core lifetime and increased thermal efficiency. Proposed ex-vessel cooling system consisted of 4 trains of thermosyphons and designed to remove 1% of thermal power with 10% of margin. The scopes of this study were design of proposed passive DHRS, validation of system analysis and optimization of system design. Mercury was selected as working fluid to design ex-vessel thermosyphon in consideration of system geometry, operating temperature and required heat flux. SUS 316 with chrome coated liner was selected as case material to resist against high corrosivity of mercury. Thermosyphon evaporator was covered on the surface of reactor vessel as the geometry of hollow shell filled with mercury. Condenser was consisted of finned tube bundles and was located in isolated water pool, the ultimate heat sink. Operation limits and thermal resistance was estimated to guarantee whether the design was adequate. System analysis was conducted by in

Endovascular treatment for acute ischaemic stroke with large vessel occlusion: the experience of a regional stroke service

International Nuclear Information System (INIS)

McCusker, M.W.; Robinson, S.; Looby, S.; Power, S.; Ti, J.P.; Grech, R.; Galvin, L.; O'Hare, A.; Brennan, P.; O'Kelly, P.; O'Brien, P.; Collins, R.; Dolan, E.; Williams, D.J.; Thornton, J.

2015-01-01

Aim: To report the experience of a regional stroke referral service with endovascular treatment for patients with acute ischaemic stroke (AIS) and large vessel occlusion. Materials and methods: A prospective review was undertaken of 93 consecutive cases receiving endovascular treatment for AIS over a 42-month period (January 2010 to June 2013). The National Institutes of Health Stroke Scale (NIHSS), location of large vessel occlusion, details of endovascular procedure, and degree of reperfusion achieved (Thrombolysis In Cerebral Infarction [TICI] score) were recorded. Mortality and functional outcome (modified Rankin Scale [mRS]) were measured at 90 days. Results: The mean patient age was 62 years (range 26–87 years). The mean NIHSS at presentation was 16 (range 6–29). All patients had confirmed proximal large-artery occlusion on computed tomography (CT) angiography: 87 in the anterior circulation, six in the posterior circulation. Of the 93 patients treated, 64 (69%) received intravenous thrombolysis. Successful reperfusion (TICI grade 2a to 3) was achieved in 80 (86%) cases. There were 13 (14%) cases of failed vessel recanalisation (TICI grade 0). Good functional outcome (mRS ≤2) was achieved in 51 (55%) cases. The 90-day mortality was 20 (22%) cases. Fifty-seven (61%) cases were transferred from outside centres. There was no significant increase in morbidity or mortality for transferred patients. Conclusion: Successful endovascular recanalisation can result in good functional outcomes for patients with AIS and large vessel occlusion. Our interventional neuroradiology service provides endovascular treatment as part of a regional stroke service without increase in morbidity or mortality for patients transferred from outside institutions. - Highlights: • Acute stoke patients may benefit from transfer to a specialist centre for endovascular treatment. • The authors offer endovascular treatment for suitable patients as part of a regional stroke service. �

Vessel Operator System

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Operator cards are required for any operator of a charter/party boat and or a commercial vessel (including carrier and processor vessels) issued a vessel permit from...

Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 1. Comprehending the vacuum vessel structure

International Nuclear Information System (INIS)

Onozuka, Masanori; Nakahira, Masataka

2006-01-01

The functions, conditions and structure of vacuum vessel using tokamak fusion machines are explained. The structural standard and code of vacuum vessel, process of vacuum vessel design, and design of ITER vacuum vessel are described. Production and maintenance of ultra high vacuum, confinement of radioactive materials, support of machines in vessel and electromagnetic force, radiation shield, plasma vertical stability, one-turn electric resistance, high temperature baking heat and remove of nuclear heat, reduce of troidal ripple, structural standard, features of safety of nuclear fusion machines, subjects of structural standard of fusion vacuum vessel, design flow of vacuum vessel, establishment of radial build, selections of materials, baking and cooling method, basic structure, structure of special parts, shield structure, and of support structure, and example of design of structure, ITER, are stated. (S.Y.)

Leak detector for reactor pressure vessel

International Nuclear Information System (INIS)

Morimoto, Mikio.

1991-01-01

A branched pipe is disposed to a leak off pipeline led from a flange surface which connects the main body and the upper lid of a reactor pressure vessel. An exhaust pump is disposed to the branched pipe and a moisture gage is disposed on the side of the exhaustion and a dry air supplier is connected to the branched pipe. Upon conducting a pressure-proof leak test for the reactor pressure vessel, the exhaust pump is operated and an electromagnet valve disposed at the upstream of the dry air supplier is opened and closed repeatedly. The humidity of air sucked by the exhaust pump is detected by the moisture gage. If leaks should be caused in the joining surface of the flange, leaked water is diffused as steams. Accordingly, occurrence of leak can be detected instantly based on the comparison with the moisture level of the dry air as a standard. In this way, a leak test can be conducted reliably in a short period of time with no change of for the reactor pressure container itself. (I.N.)

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

Ultrasonic examination of defects close to the outer surface

International Nuclear Information System (INIS)

Benoist, P.; Serre, M.; Champigny, F.

1986-11-01

During the examination of a pressurized water reactor vessel with an in Service Inspection Machine (MIS), various welds are scanned with immersion ultrasonic focused transducers from the inside of the vessel. Defects close to the outer surface are sometimes detected, and sizing with the successive 6 dB drop method leads to oversize some indications; this is caused by various reflections on the outer wall; the corner echo is of particular importance here. CEA and EDF have started an experimental program in order to study the response of volumetric and planar defects located near the outer surface. We present here the first results obtained with artificial defects. 2 refs

Trigeminal neuralgia: how often are trigeminal nerve-vessel contacts found by MRI in normal volunteers

International Nuclear Information System (INIS)

Kress, B.; Schindler, M.; Haehnel, S.; Sartor, K.; Rasche, D.; Tronnier, V.

2006-01-01

Purpose: To assess prospectively how often contacts are found between the trigeminal nerve and arteries or veins in the perimesencephalic cistern via MRI in normal volunteers. Materials and methods: 48 volunteers without a history of trigeminal neuralgia were examined prospectively (MRI at 1.5T; T2-CISS sequence, coronal orientation, 0.9 mm slice thickness). Two radiologists decided by consensus whether there was a nerve-vessel contact in the perimesencephalic cistern. Results: In 27% of the volunteers, no contact was found between the trigeminal nerve and regional vessels, while in 73%, such a contact was present. In 61% of the cases, the offending vessel was an artery, in 39%, it was a vein. In 2 volunteers, a deformation of the nerve was noted. Conclusion: Contrary to what has been suggested by retrospective studies, the majority of normal volunteers, if studied prospectively, do show a contact between the trigeminal nerve and local vessels. A close proximity between the nerve and regional vessels is thus normal and is not necessarily proof of a pathological nerve-vessel conflict. (orig.)

Quality assurance experience in the manufacture of PFBR reactor vessel during technology development work

International Nuclear Information System (INIS)

Shanmugam, K.; Chandramohan, R.; Ramamurthy, M.K.

1996-01-01

An efficient and proper implementation of quality assurance in the technology development works of Prototype Fast Breeder Reactor (PFBR) main vessel was undertaken to achieve the desired quality and dimensional accuracy of main vessel. In this paper an attempt has been made to bring out the methods and procedures adopted to implement the quality assurance programme on important activities including approval of documents, material, general requirements for manufacture of SS components, inspection procedures, forming and welding of petals, non-destructive testing etc. (author)

Prestressed concrete vessels suitable for helium high temperature reactors

International Nuclear Information System (INIS)

Lockett, G.E.; Kinkead, A.N.

1967-02-01

In considering prestressed concrete vessels for use with helium cooled high temperature reactors, a number of new problems arise and projected designs involve new approaches and new solutions. These reactors, having high coolant outlet temperature from the core and relatively high power densities, can be built into compact designs which permit usefully high working pressures. Consequently, steam generators and circulating units tend to be small. Although circuit activity can be kept quite low with coated particle fuels, designs which involve entry for subsequent repair are not favoured, and coupled with the preferred aim of using fully shop fabricated units within the designs with removable steam generators which involve no tube welding inside the vessel. A particular solution uses a number of slim cylindrical assemblies housed in the wall of the pressure vessel and this vessel design concept is presented. The use of helium requires very high sealing standards and one of the important requirements is a vessel design which permits leak testing during construction, so that a repair seal can be made to any faulty part in a liner seam. Very good demountable joint seals can be made without particular difficulty and Dragon experience is used to provide solutions which are suitable for prestressed concrete vessel penetrations. The concept layout is given of a vessel meeting these requirements; the basis of design is outlined and special features of importance discussed. (author)

Effects of the vessel core seismic interaction for a fast reactor

International Nuclear Information System (INIS)

Martelli, A.; Maresca, G.

1983-01-01

This report describes the features and the results of the analysis carried out in collaboration between ENEA and NIRA for evaluating the effects of the vessel core dynamic interaction in case of safe shutdown earthquake, both in absence and in presence of one or two restraint plates inserted in the tank close to the middle and or top planes for limiting the core seismic motion. Such analysis, although carried out making use of preliminary data, contributed to the recent ENEA decision of applying a core restraint close to the core element top

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

First results of closed helical divertor experiment in LHD

International Nuclear Information System (INIS)

Morisaki, T.; Masuzaki, S.; Kobayashi, M.

2012-11-01

The baffle-structured closed Helical Divertor (CHD) is being constructed in LHD to actively control the edge plasma, which consists of ten discrete modules installed on inboard side of the torus. At this stage, two of ten modules have been constructed. In the initial experiments, performance of CHD was experimentally investigated, comparing with numerical expectations. During the continuous gas puffing discharge, it was observed the neutral pressure in the CHD was more than 10 times higher than that in the open HD, which agrees well with the numerical simulation. In the high density regime, indication of the divertor detachment was observed in CHD, which was caused by the high recycling and high density state in CHD. With a Penning discharge diagnostics, the neutral particle behaviour with different species was investigated. Little difference between hydrogen and helium was observed in transport property. (author)

Development efforts on helium vessel for 5 cell - 650 MHz SRF cavity at RRCAT

International Nuclear Information System (INIS)

Kumar, Abhay; Kumar, Pankaj; Sandha, R.S.; Dutta, Subhajit; Soni, Rakesh; Dwivedi, Jishnu; Thakurta, A.C.; Bhatnagar, V.K.; Mundra, G.

2011-01-01

The work focuses on the development of helium vessel which houses a 5 cell - 650 MHz SRF niobium cavity and serves as a helium bath to maintain the cavity at 2 K. The vessel has provision for changing the axial length of the cavity for tuning purpose by using a tuning mechanism and a large bellow. Titanium has been chosen as a material of construction of the vessel due to its coefficient of thermal expansion being close to that of niobium. Efforts have been initiated to understand the functional requirements, design requirements, acceptance criteria for design and analysis, non-destructive examination requirements, inspection and testing requirements, manufacturing technology of the titanium vessel and its integration with the SRF cavity. The welding assumes a special significance as titanium is highly reactive and ductility of the weld joint is lost in the presence of air and other impurities. A trial vessel has been conceptualised having typical sizes and geometries. The manufacturing features of vessel are based on ASME B and PV Code, Section VIII Division-1 and manufacturing of this vessel has been started at an Indian industry. Quality assurance plan for this work is developed. The paper describes the work done at RRCAT on the functional and integration requirements, overall design requirements, design methodology to achieve code conformance, manufacturing technology and QAP being used in the development of helium vessel. (author)

Automated method for identification and artery-venous classification of vessel trees in retinal vessel networks.

Science.gov (United States)

Joshi, Vinayak S; Reinhardt, Joseph M; Garvin, Mona K; Abramoff, Michael D

2014-01-01

The separation of the retinal vessel network into distinct arterial and venous vessel trees is of high interest. We propose an automated method for identification and separation of retinal vessel trees in a retinal color image by converting a vessel segmentation image into a vessel segment map and identifying the individual vessel trees by graph search. Orientation, width, and intensity of each vessel segment are utilized to find the optimal graph of vessel segments. The separated vessel trees are labeled as primary vessel or branches. We utilize the separated vessel trees for arterial-venous (AV) classification, based on the color properties of the vessels in each tree graph. We applied our approach to a dataset of 50 fundus images from 50 subjects. The proposed method resulted in an accuracy of 91.44% correctly classified vessel pixels as either artery or vein. The accuracy of correctly classified major vessel segments was 96.42%.

Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

Science.gov (United States)

Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

A numerical analysis on the curved bileaflet Mechanical Heart Valve (MHV) : leaflet motion and blood flow in an elastic blood vessel

International Nuclear Information System (INIS)

Bang, Jin Seok; Kim, Chang Nyung; Choi, Choeng Ryul

2005-01-01

In blood flow passing through the Mechanical Heart Valve (MHV) and elastic blood vessel, hemolysis and platelet activation causing thrombus formation can be seen owing to the shear stress in the blood. Also, fracture and deformation of leaflets can be observed depending on the shape and material properties of the leaflets which is opened and closed in a cycle. Hence, comprehensive study is needed on the hemodynamics which is associated with the motion of leaflet and elastic blood vessel in terms of fluid-structure interaction. In this paper, a numerical analysis has been performed for a three-dimensional pulsatile blood flow associated with the elastic blood vessel and curved bileaflet for multiple cycles in light of fluid-structure interaction. From this analysis fluttering phenomenon and rebound of the leaflet have been observed and recirculation and regurgitation have been found in the flow fields of the blood. Also, the pressure distribution and the radial displacement of the elastic blood vessel have been obtained. The motion of the leaflet and flow fields of the blood have shown similar tendency compared with the previous experiments carried out in other studies. The present study can contribute to the design methodology for the curved bileaflet mechanical heart valve. Furthermore, the proposed fluid-structure interaction method will be effectively used in various fields where the interaction between fluid flow and structure are involved

Installation method for the steel container and vessel of the nuclear heating reactor

International Nuclear Information System (INIS)

Chen Liying; Guo Jilin; Liu Wei

2000-01-01

The Nuclear Heating Reactor (NHR) has the advantages of inherent safety and better economics, integrated arrangement, full power natural circulation and dual vessel structure. However, the large thin container presents a new and difficult problem. The characteristics of the dual vessel installation method are analyzed with system engineering theory. Since there is no foreign or domestic experience, a new method was developed for the dual vessel installation for the 5 MW NHR. The result shows that the installation method is safe and reliable. The research on the dual vessel installation method has important significance for the design, manufacture and installation of the NHR dual vessel, as well as the industrialization and standardization of the NHR

Pressure vessel for nuclear reactor plant consisting of several pre-stressed cast pressure vessels

International Nuclear Information System (INIS)

Bodmann, E.

1984-01-01

Several cylindrical pressure vessel components made of pressure castings are arranged on a sector of a circle around the cylindrical cast pressure vessel for accommodating the helium cooled HTR. Each component pressure vessel is connected to the reactor vessel by a horizontal gas duct. The contact surfaces between reactor and component pressure vessel are in one plane. In the spaces between the individual component pressure vessels, there are supporting blocks made of cast iron, which are hollow and also have flat surfaces. With the reactor vessel and the component pressure vessels they form a disc-shaped connecting part below and above the gas ducts. (orig./PW)

Exploring the Use of Design of Experiments in Industrial Processes Operating Under Closed-Loop Control

DEFF Research Database (Denmark)

Capaci, Francesca; Kulahci, Murat; Vanhatalo, Erik

2017-01-01

Industrial manufacturing processes often operate under closed-loop control, where automation aims to keep important process variables at their set-points. In process industries such as pulp, paper, chemical and steel plants, it is often hard to find production processes operating in open loop....... Instead, closed-loop control systems will actively attempt to minimize the impact of process disturbances. However, we argue that an implicit assumption in most experimental investigations is that the studied system is open loop, allowing the experimental factors to freely affect the important system...... responses. This scenario is typically not found in process industries. The purpose of this article is therefore to explore issues of experimental design and analysis in processes operating under closed-loop control and to illustrate how Design of Experiments can help in improving and optimizing...

Mass transfer experiments for the heat load during in-vessel retention of core melt

Energy Technology Data Exchange (ETDEWEB)

Park, Hae Kyun; Chung, Bum Jin [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

2016-08-15

We investigated the heat load imposed on the lower head of a reactor vessel by the natural convection of the oxide pool in a severe accident. Mass transfer experiments using a CuSO{sub 4}–H{sub 2}SO{sub 4} electroplating system were performed based on the analogy between heat and mass transfer. The Ra′{sub H} of 10{sup 14} order was achieved with a facility height of only 0.1 m. Three different volumetric heat sources were compared; two had identical configurations to those previously reported, and the other was designed by the authors. The measured Nu's of the lower head were about 30% lower than those previously reported. The measured angular heat flux ratios were similar to those reported in existing studies except for the peaks appearing near the top. The volumetric heat sources did not affect the Nu of the lower head but affected the Nu of the top plate by obstructing the rising flow from the bottom.

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.

ITER vacuum vessel design and electromagnetic analysis on in-vessel components

International Nuclear Information System (INIS)

Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.; Iizuka, T.

1995-01-01

Major functional requirements for the vacuum vessel are to provide the first safety barrier and to support electromagnetic loads due to plasma disruptions and vertical displacement events, and to withstand plausible accidents without losing confinement. A double wall structure concept has been developed for the vacuum vessel due to its beneficial characteristics from the viewpoints of structural integrity and electrical continuity. An electromagnetic analysis of the blanket modules and the vacuum vessel has been performed to investigate force distributions on in-vessel components. According to the vertical displacement events (VDE) scenario, which assumes a critical q-value of 1.5, the total downward vertical force, induced by coupling between the eddy current and external fields, is about 110 MN. We have performed a stress analysis for the vacuum vessel using the VDE disruption forces acting on the blankets, and a maximum stress intensity of 112 MPa was obtained in the vicinity of the lower support of the vessel. (orig.)

An automated vessel segmentation of retinal images using multiscale vesselness

International Nuclear Information System (INIS)

Ben Abdallah, M.; Malek, J.; Tourki, R.; Krissian, K.

2011-01-01

The ocular fundus image can provide information on pathological changes caused by local ocular diseases and early signs of certain systemic diseases, such as diabetes and hypertension. Automated analysis and interpretation of fundus images has become a necessary and important diagnostic procedure in ophthalmology. The extraction of blood vessels from retinal images is an important and challenging task in medical analysis and diagnosis. In this paper, we introduce an implementation of the anisotropic diffusion which allows reducing the noise and better preserving small structures like vessels in 2D images. A vessel detection filter, based on a multi-scale vesselness function, is then applied to enhance vascular structures.

Factors affecting the integrity of PWR pressure vessels during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.

1983-01-01

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, if certain postulated accidents, referred to as overcooling accidents, were to occur, the pressure vessel could be subjected to severe thermal shock while the pressure is substantial. As a result, 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 fracture-mechanics analysis for a typical postulated accident and also related thermal-shock experiments indicate that very shallow surface flaws that extend through the cladding into the base material could propagate. This is of particular concern because shallow flaws appear to be the most probable and presumably are the most difficult to detect

Decommissioning experience of the Japan power demonstration reactor

International Nuclear Information System (INIS)

Hoshi, T.; Yanagihara, S.; Tachibana, M.; Momma, T.

1992-01-01

Actual dismantling of the Japan Power Demonstration Reactor (JPDR) has been progressing since 1986 aiming to make stage 3 condition as the final goal. Such highly activated components as the reactor pressure vessel (RPV) and the inner portion of biological shield concrete close to the RPV have removed using the remotely operated cutting machines. Useful data on the dismantling techniques and their safety as well as the manpower expenditure and radiation exposure of workers have been obtained. Experiences gained through the dismantling works are described in this paper. (author)

Research vessels

Digital Repository Service at National Institute of Oceanography (India)

Rao, P.S.

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

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

Inductive testing of reactor pressure vessels

International Nuclear Information System (INIS)

Bergh, H.

1987-01-01

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

Study on severe fuel damage and in-vessel melt progression

International Nuclear Information System (INIS)

Kim, Hee Dong; Kim, Sang Baik; Lee, Gyu Jung

1992-06-01

In-vessel core melt progression describes the progression of the state of a reactor core from core uncovery up to reactor vessel melt through in uncovered accidents or through temperature stabilization in accidents recovered by core reflooding. Melt progression can be thought as two parts; early melt progression and late melt progression. Early phase of core melt progression includes the progression of core material melting and relocation, which mostly consist of metallic materials. On the other hand, the late phase of core melt progression involves ceramic material melt and relocation to the lower plenum and heat-up the reactor vessel lower head. A large number of information are available for the early melt progression through experiments such as SFD, DF, FLHT test and utilized in the severe accident analysis codes. However, understanding of the late phase melt progression phenomenology is based primary on TMI-2 core examinations and not much experimental information is available. Especilally, the great uncertainties exist in vessel failure mode, melt composition, mass, and temperature. Further research is planned to perform to reduce the uncertainties in understanding of core melt down accidents as parts of long term melt progression research program. A study on the core melt progression at KAERI has been being performed through the Severe Accident Research Program with USNRC. KAERI staff had participated in the PBF SFD experiments at INEL and analyses of experiments were performed using SCDAP code. Experiments of core melt program have not been carried out at KAERI yet. It is planned that further research on core melt down accidents will be performed, which is related to design of future generations of nuclear reactors as parts of long-term project for improvement of nuclear reactor safety. (Author)

Cavitation damage prediction for the JSNS mercury target vessel

Energy Technology Data Exchange (ETDEWEB)

Naoe, Takashi, E-mail: naoe.takashi@jaea.go.jp; Kogawa, Hiroyuki; Wakui, Takashi; Haga, Katsuhiro; Teshigawara, Makoto; Kinoshita, Hidetaka; Takada, Hiroshi; Futakawa, Masatoshi

2016-01-15

The liquid mercury target system for the Japan Spallation Neutron Source (JSNS) at the Materials and Life science experimental Facility (MLF) in the Japan Proton Accelerator Research Complex (J-PARC) is designed to produce pulsed neutrons. The mercury target vessel in this system, which is made of type 316L stainless steel, is damaged by pressure wave-induced cavitation due to proton beam bombardment. Currently, cavitation damage is considered to be the dominant factor influencing the service life of the target vessel rather than radiation damage. In this study, cavitation damage to the interior surface of the target vessel was predicted on the basis of accumulated damage data from off-beam and on-beam experiments. The predicted damage was compared with the damage observed in a used target vessel. Furthermore, the effect of injecting gas microbubbles on cavitation damage was predicted through the measurement of the acoustic vibration of the target vessel. It was shown that the predicted depth of cavitation damage is reasonably coincident with the observed results. Moreover, it was confirmed that the injection of gas microbubbles had an effect on cavitation damage.

Structural considerations in design of lightweight glass-fiber composite pressure vessels

Science.gov (United States)

Faddoul, J. R.

1973-01-01

The design concepts used for metal-lined glass-fiber composite pressure vessels are described, comparing the structural characteristics of the composite designs with each other and with homogeneous metal pressure vessels. Specific design techniques and available design data are identified. The discussion centers around two distinctly different design concepts, which provide the basis for defining metal lined composite vessels as either (1) thin-metal lined, or (2) glass fiber reinforced (GFR). Both concepts are described and associated development problems are identified and discussed. Relevant fabrication and testing experience from a series of NASA-Lewis Research Center development efforts is presented.

36 CFR 13.1178 - Closed waters, islands and other areas.

Science.gov (United States)

2010-07-01

... southeast of Flapjack Island; or Eider Island; or Boulder Island; or Geikie Rock; or Lone Island; or the... islands) of the easternmost point of Russell Island; or Graves Rocks (on the outer coast); or Cormorant... and Preserve Vessel Operating Restrictions § 13.1178 Closed waters, islands and other areas. The...

Integration of ITER in-vessel diagnostic components in the vacuum vessel

International Nuclear Information System (INIS)

Encheva, A.; Bertalot, L.; Macklin, B.; Vayakis, G.; Walker, C.

2009-01-01

The integration of ITER in-vessel diagnostic components is an important engineering activity. The positioning of the diagnostic components must correlate not only with their functional specifications but also with the design of the major parts of ITER torus, in particular the vacuum vessel, blanket modules, blanket manifolds, divertor, and port plugs, some of which are not yet finally designed. Moreover, the recently introduced Edge Localised Mode (ELM)/Vertical Stability (VS) coils mounted on the vacuum vessel inner wall call for not only more than a simple review of the engineering design settled down for several years now, but also for a change in the in-vessel distribution of the diagnostic components and their full impact has yet to be determined. Meanwhile, the procurement arrangement (a document defining roles and responsibilities of ITER Organization and Domestic Agency(s) (DAs) for each in-kind procurement including technical scope of work, quality assurance requirements, schedule, administrative matters) for the vacuum vessel must be finalized. These make the interface process even more challenging in terms of meeting the vacuum vessel (VV) procurement arrangement's deadline. The process of planning the installation of all the ITER diagnostics and integrating their installation into the ITER Integrated Project Schedule (IPS) is now underway. This paper covers the progress made recently on updating and issuing the interfaces of the in-vessel diagnostic components with the vacuum vessel, outlines the requirements for their attachment and summarises the installation sequence.

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

Recovery of testicular blood flow following ligation of testicular vessels

International Nuclear Information System (INIS)

Pascual, J.A.; Villanueva-Meyer, J.; Salido, E.; Ehrlich, R.M.; Mena, I.; Rajfer, J.

1989-01-01

To determine whether initial ligation of the testicular vessels of the high undescended testis followed by a delayed secondary orchiopexy is a viable alternative to the classical Fowler-Stephens procedure, a series of preliminary experiments were conducted in the rat in which testicular blood flow was measured by the 133-xenon washout technique before, and 1 hour and 30 days after ligation of the vessels. In addition, testicular histology, and testis and sex-accessory tissue weights were measured in 6 control, 6 sham operated and 6 testicular vessel ligated rats 54 days after vessel ligation. The data demonstrate that ligation and division of the testicular blood vessels produce an 80 per cent decrease in testicular blood flow 1 hour after ligation of the vessels. However, 30 days later testis blood flow returns to the control and pre-treatment value. There were no significant changes in testis or sex-accessory tissue weights 54 days after vessel ligation. Histologically, 4 of the surgically operated testes demonstrated necrosis of less than 25 per cent of the seminiferous tubules while 1 testis demonstrated more than 75 per cent necrosis. The rest of the tubules in all 6 testes demonstrated normal spermatogenesis. From this study we conclude that initial testicular vessel ligation produces an immediate decrease in testicular blood flow but with time the collateral vessels are able to compensate and return the testis blood flow to its normal pre-treatment value. These preliminary observations lend support for the concept that initial ligation of the testicular vessels followed by a delayed secondary orchiopexy in patients with a high undescended testis may be a possible alternative to the classical Fowler-Stephens approach

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

Primary care closed claims experience of Massachusetts malpractice insurers.

Science.gov (United States)

Schiff, Gordon D; Puopolo, Ann Louise; Huben-Kearney, Anne; Yu, Winnie; Keohane, Carol; McDonough, Peggy; Ellis, Bonnie R; Bates, David W; Biondolillo, Madeleine

Despite prior focus on high-impact inpatient cases, there are increasing data and awareness that malpractice in the outpatient setting, particularly in primary care, is a leading contributor to malpractice risk and claims. To study patterns of primary care malpractice types, causes, and outcomes as part of a Massachusetts ambulatory malpractice risk and safety improvement project. Retrospective review of pooled closed claims data of 2 malpractice carriers covering most Massachusetts physicians during a 5-year period (January 1, 2005, through December 31, 2009). Data were harmonized between the 2 insurers using a standardized taxonomy. Primary care practices in Massachusetts. All malpractice claims that involved primary care practices insured by the 2 largest insurers in the state were screened. A total of 551 claims from primary care practices were identified for the analysis. Numbers and types of claims, including whether claims involved primary care physicians or practices; classification of alleged malpractice (eg, misdiagnosis or medication error); patient diagnosis; breakdown in care process; and claim outcome (dismissed, settled, verdict for plaintiff, or verdict for defendant). During a 5-year period there were 7224 malpractice claims of which 551 (7.7%) were from primary care practices. Allegations were related to diagnosis in 397 (72.1%), medications in 68 (12.3%), other medical treatment in 41 (7.4%), communication in 15 (2.7%), patient rights in 11 (2.0%), and patient safety or security in 8 (1.5%). Leading diagnoses were cancer (n = 190), heart diseases (n = 43), blood vessel diseases (n = 27), infections (n = 22), and stroke (n = 16). Primary care cases were significantly more likely to be settled (35.2% vs 20.5%) or result in a verdict for the plaintiff (1.6% vs 0.9%) compared with non-general medical malpractice claims (P < .001). In Massachusetts, most primary care claims filed are related to alleged misdiagnosis. Compared with malpractice

Intracranial arterial aneurysm vasculopathies: targeting the outer vessel wall

International Nuclear Information System (INIS)

Krings, Timo; Piske, Ronie L.; Lasjaunias, Pierre L.

2005-01-01

The pathogenesis of intracranial arterial aneurysms (AA) remains unclear, despite their clinical importance. An improved understanding of this disease is important in choosing therapeutic options. In addition to the ''classical'' berry-type aneurysm, there are various other types of intracranial AA such as infectious, dissecting or giant, partially-thrombosed aneurysms. From the clinician's perspective, the hypothesis that some of these intracranial AA might be due to abluminal factors has been proposed for several years. Indeed, this hypothesis and the empirical use of anti-inflammatory drugs in giant intracranial aneurysms have been confirmed by recent studies reporting that an enzyme involved in the inflammatory cascade (5-lipoxygenase or 5-LO) promotes the pathogenesis of specific aneurysms in humans. 5-LO generates different forms of leukotrienes which are potent mediators of inflammation. Adventitial inflammation leads to a weakening of the media from the abluminal part of the vessel wall due to the release of proinflammatory factors that invade the media, thereby degrading the extracellular matrix, the elastic lamina of the vascular wall, and, finally, the integrity of the vessel lumen. This in turn results in a dilation of the vessel and aneurysm formation. Moreover, neoangiogenesis of vasa vasorum is found in close proximity to 5-LO activated macrophages. In addition to this biological cascade, we argue that repeated subadventitial haemorrhages from the new vasa vasorum play an important role in aneurysm pathogenesis, due to a progressive increase in size mediated by the apposition of new layers of intramural haematoma within the vessel wall. Intracranial giant AA can therefore be regarded as a proliferative disease of the vessel wall induced by extravascular activity. (orig.)

Concrete containment vessels (CCV) for nuclear power plants, (1)

International Nuclear Information System (INIS)

Ibe, Yukimi; Kitajima, Masatake

1977-01-01

Containment vessels (CV) and the construction of concrete containment vessels (CCV) for nuclear power plants are described generally, and their use and techniques in foreign countries are illustrated, in connection with the introduction of CCV to Japanese nuclear power plants. The introduction deals with the construction plan of Japanese nuclear power plants, and with the difficulties in the steel CV for large scale construction. The investigations, tests and researches are not yet sufficient. The prompt establishment of safety supported by technical criteria, analytical methods and experiments is desired. The second part deals with the consideration for aseismatic design, construction, function and characteristics of CCV. The classification and currently employed CCV, which is mainly reinforced concrete containment vessels (RCCV), are described, and the typical CCV employed for BWR is illustrated. Further, the typical arrangement of reinforcing steels at the cylindrical portion and the dome portion of RCCV is illustrated. The third part deals with the present state of CCV abroad. A prestressed concrete containment vessel (PCCV) of Turkey Point power plant is illustrated as a typical example of CCV. The tests reported in the international meeting for the design, construction and operation of concrete pressure vessels and concrete containment vessels at York University in England in 1975 are reviewed. Typical examples of the design conditions, the size and form, and the construction procedure for PCCV and RCCV abroad are reviewed. (Iwakiri, K.)

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)

Thermal performance of an insulating structure for a reactor vessel

International Nuclear Information System (INIS)

Aranovitch, E.; Crutzen, S.; LeDet, M.; Denis, R.

This report describes the installations used to test the HTGR reactor vessel insulating structure called ''Casali'' and details the experimental results in 3 groups: general experiments, systematic study, and technological experiments. The results obtained make it possible to satisfactorily predict the behavior of the structure in a practical application

Autonomous Radiation Monitoring of Small Vessels

International Nuclear Information System (INIS)

Fabris, Lorenzo; Hornback, Donald Eric

2010-01-01

Small private vessels are one avenue by which nuclear materials may be smuggled across international borders. While one can contemplate using the terrestrial approach of radiation portal monitors on the navigable waterways that lead to many ports, these systems are ill-suited to the problem. They require vehicles to pass at slow speeds between two closely-spaced radiation sensors, relying on the uniformity of vehicle sizes to space the detectors, and on proximity to link an individual vehicle to its radiation signature. In contrast to roadways where lanes segregate vehicles, and motion is well controlled by inspection booths; channels, inlets, and rivers present chaotic traffic patterns populated by vessels of all sizes. We have developed a unique solution to this problem based on our portal-less portal monitor instrument that is designed to handle free-flowing traffic on roadways with up to five-traffic lanes. The instrument uses a combination of visible-light and gamma-ray imaging to acquire and link radiation images to individual vehicles. It was recently tested in a maritime setting. In this paper we present the instrument, how it functions, and the results of the recent tests.

Minilaparoscopic varicocelectomy with preservation of testicular artery and lymphatic vessels by using intracorporeal knot-tying technique: five-year experience.

Science.gov (United States)

Chung, Shiu-Dong; Wu, Chia-Chang; Lin, Victor Chia-Hsiang; Ho, Chen-Hsun; Yang, Stephen Shei Dei; Tsai, Yao-Chou

2011-08-01

In this study we present our experience using minilaparoscopic intracorporeal knot tying to ligate internal spermatic veins (ISV) while sparing the spermatic artery and lymphatics. Minilaparoscopic varicocelectomies were performed in 87 patients between January 2004 and January 2009. All varicoceles were detected clinically according to the World Health Organization (WHO) classification and confirmed by scrotal color Doppler ultrasonography. The surgical indications were scrotal symptoms in 71, infertility in 16, and both conditions in 2. Three 3.5 mm minilaparoscopic ports were used for the operation. The ISVs were dissected and then ligated with intracorporeal knot-tying. The testicular artery and lymphatic vessels were carefully preserved to minimize procedure-related complications. Unilateral laparoscopic varicocelectomy was performed in 21 (24.2%) patients and bilateral in 66 (75.8%). Mean operative time was 71.1 ± 29.2 and 46.8 ± 12.6 min for bilateral and unilateral varicocelectomies, respectively. All patients were discharged within 24 h after surgery. Neither immediate major nor late procedure-related complications were noted. Of the 71 patients with scrotal symptoms, the symptoms completely subsided in 55 (77.5%) and partially subsided in 10 (14.1%). Only one (1.2%) recurrent varicocele was detected within a mean follow-up of 21 months (range = 3-42). Neither hydrocele formation nor testicular atrophy was found during the follow-up period. Our 5-year experience revealed that minilaparoscopic varicocelectomy with sparing of artery and lymphatic vessels could safely and effectively ligate all spermatic veins and preserve spermatic arteries and lymphatic channels without leading to a high varicocele persistence or recurrence.

Closing the Gap GEF Experiences in Global Energy Efficiency

CERN Document Server

Yang, Ming

2013-01-01

Energy efficiency plays and will continue to play an important role in the world to save energy and mitigate greenhouse gas (GHG) emissions. However, little is known on how much additional capital should be invested to ensure using energy efficiently as it should be, and very little is known which sub-areas, technologies, and countries shall achieve maximum greenhouse gas emissions mitigation per dollar of investment in energy efficiency worldwide. Analyzing completed and slowly moving energy efficiency projects by the Global Environment Facility during 1991-2010, Closing the Gap: GEF Experiences in Global Energy Efficiency evaluates impacts of multi-billion-dollar investments in the world energy efficiency. It covers the following areas: 1.       Reviewing the world energy efficiency investment and disclosing the global energy efficiency gap and market barriers that cause the gap; 2.       Leveraging private funds with public funds and other resources in energy efficiency investments; using...

Research on release rate of volatile organic compounds in typical vessel cabin

Directory of Open Access Journals (Sweden)

ZHANG Jinlan

2018-02-01

Full Text Available [Objectives] Volatile Organic Compounds (VOC should be efficiently controlled in vessel cabins to ensure the crew's health and navigation safety. As an important parameter, research on release rate of VOCs in cabins is required. [Methods] This paper develops a method to investigate this parameter of a ship's cabin based on methods used in other closed indoor environments. A typical vessel cabin is sampled with Tenax TA tubes and analyzed by Automated Thermal Desorption-Gas Chromatography-Mass Spectrometry (ATD-GC/MS. The lumped mode is used and the release rate of Benzene, Toluene, Ethylbenzene and Xylene (BTEX, the typical representatives of VOCs, is obtained both in closed and ventilated conditions. [Results] The results show that the content of xylene and Total Volatile Organic Compounds (TVOC exceed the indoor environment standards in ventilated conditions. The BTEX release rate is similar in both conditions except for the benzene. [Conclusions] This research builds a method to measure the release rate of VOCs, providing references for pollution character evaluation and ventilation and purification system design.

Latest feedback from a major reactor vessel dismantling project

International Nuclear Information System (INIS)

Boucau, J.; Segerud, P.; Sanchez, M.; Garcia, R.

2015-01-01

Westinghouse performed two large segmentation projects in 2010-2013 and then 2013-2015 at the Jose Cabrera nuclear power plant in Spain. The power plant is located in Almonacid de Zorita, 43 miles east of Madrid, Spain and was in operation between 1968 and 2006. This paper will describe the sequential steps required to prepare, segment, separate, and package the individual component segments using under water mechanical techniques. The paper will also include experiences and lessons learned that Westinghouse has collected from the activities performed during the reactor vessel and vessel internals segmentation projects. (authors)

Examination of VVER-1000 Reactor Pressure Vessel

International Nuclear Information System (INIS)

Matokovic, A.; Picek, E.; Markulin, K.

2008-01-01

The increasing demand of a higher level of safety in the operation of the nuclear power plants requires the utilisation of more precise automated equipment to perform in-service inspections. That has been achieved by technological advances in computer technology, in robotics, in examination probe technology with the development of the advanced inspection technique and has also been due to the considerable and varied experience gained in the performance of such inspections. In-service inspection of reactor pressure vessel, especially Russian-designed WWER-1000 presents one of the most important and extensive examination of nuclear power plants primary circuit components. Such examination demand high standards of inspection technology, quality and continual innovation in the field of non-destructive testing advanced technology. A remote underwater contact ultrasonic technique is employed for the examination of the base metal of vessel and reactor welds, whence eddy current method is applied for clad surface examinations. Visual testing is used for examination of the vessel interior. The movement of inspection probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with inspection systems. The successful performance of reactor pressure vessel is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen non-destructive techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state-of-the-art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. This paper presents advanced approach in the reactor pressure vessel in-service inspections and it is especially developed for WWER-1000 nuclear power plants.(author)

Intravital two-photon microscopy of immune cell dynamics in corneal lymphatic vessels.

Directory of Open Access Journals (Sweden)

Philipp Steven

Full Text Available BACKGROUND: The role of lymphatic vessels in tissue and organ transplantation as well as in tumor growth and metastasis has drawn great attention in recent years. METHODOLOGY/PRINCIPAL FINDINGS: We now developed a novel method using non-invasive two-photon microscopy to simultaneously visualize and track specifically stained lymphatic vessels and autofluorescent adjacent tissues such as collagen fibrils, blood vessels and immune cells in the mouse model of corneal neovascularization in vivo. The mouse cornea serves as an ideal tissue for this technique due to its easy accessibility and its inducible and modifiable state of pathological hem- and lymphvascularization. Neovascularization was induced by suture placement in corneas of Balb/C mice. Two weeks after treatment, lymphatic vessels were stained intravital by intrastromal injection of a fluorescently labeled LYVE-1 antibody and the corneas were evaluated in vivo by two-photon microscopy (TPM. Intravital TPM was performed at 710 nm and 826 nm excitation wavelengths to detect immunofluorescence and tissue autofluorescence using a custom made animal holder. Corneas were then harvested, fixed and analyzed by histology. Time lapse imaging demonstrated the first in vivo evidence of immune cell migration into lymphatic vessels and luminal transport of individual cells. Cells immigrated within 1-5.5 min into the vessel lumen. Mean velocities of intrastromal corneal immune cells were around 9 µm/min and therefore comparable to those of T-cells and macrophages in other mucosal surfaces. CONCLUSIONS: To our knowledge we here demonstrate for the first time the intravital real-time transmigration of immune cells into lymphatic vessels. Overall this study demonstrates the valuable use of intravital autofluorescence two-photon microscopy in the model of suture-induced corneal vascularizations to study interactions of immune and subsequently tumor cells with lymphatic vessels under close as possible

Special enclosure for a pressure vessel

International Nuclear Information System (INIS)

Wedellsborg, B.W.; Wedellsborg, U.W.

1993-01-01

A pressure vessel enclosure is described comprising a primary pressure vessel, a first pressure vessel containment assembly adapted to enclose said primary pressure vessel and be spaced apart therefrom, a first upper pressure vessel jacket adapted to enclose the upper half of said first pressure vessel containment assembly and be spaced apart therefrom, said upper pressure vessel jacket having an upper rim and a lower rim, each of said rims connected in a slidable relationship to the outer surface of said first pressure vessel containment assembly, mean for connecting in a sealable relationship said upper rim of said first upper pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, means for connecting in a sealable relationship said lower rim of said first upper pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, a first lower pressure vessel jacket adapted to enclose the lower half of said first pressure vessel containment assembly and be spaced apart therefrom, said lower pressure vessel jacket having an upper rim connected in a slidable relationship to the outer surface of said first pressure vessel containment assembly, and means for connecting in a sealable relationship said upper rim of said first lower pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, a second upper pressure vessel jacket adapted to enclose said first upper pressure vessel jacket and be spaced apart therefrom, said second upper pressure vessel jacket having an upper rim and a lower rim, each of said rims adapted to slidably engage the outer surface of said first upper pressure vessel jacket, means for sealing said rims, a second lower pressure vessel jacket adapted to enclose said first lower pressure vessel jacket and be spaced apart therefrom

Vessel classification method based on vessel behavior in the port of Rotterdam

NARCIS (Netherlands)

Zhou, Y.; Daamen, W.; Vellinga, T.; Hoogendoorn, S.P.

2015-01-01

AIS (Automatic Identification System) data have proven to be a valuable source to investigate vessel behavior. The analysis of AIS data provides a possibility to recognize vessel behavior patterns in a waterway area. Furthermore, AIS data can be used to classify vessel behavior into several

Vessel Operating Units (Vessels)

Data.gov (United States)

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

Shock loading of reactor vessel following hypothetical core disruptive accident

International Nuclear Information System (INIS)

Srinivas, G.; Doshi, J.B.

1990-01-01

Hypothetical Core Disruptive Accident (HCDA) has been historically considered as the maximum credible accident in Fast Breeder Reactor systems. Environmental consequences of such an accident depends to a great extent on the ability of the reactor vessel to maintain integrity during the shock loading following an HCDA. In the present paper, a computational model of the reactor core and the surrounding coolant with a free surface is numerical technique. The equations for conservation of mass, momentum and energy along with an equation of state are considered in two dimensional cylindrical geometry. The reactor core at the end of HCDA is taken as a bubble of hot, vaporized fuel at high temperature and pressure, formed at the center of the reactor vessel and expanding against the surrounding liquid sodium coolant. The free surface of sodium at the top of the vessel and the movement of the core bubble-liquid coolant interface are tracked by Marker and Cell (MAC) procedure. The results are obtained for the transient pressure at the vessel wall and also for the loading on the roof plug by the impact of the slug of liquid sodium. The computer code developed is validated against a benchmark experiment chosen to be ISPRA experiment reported in literature. The computer code is next applied to predict the loading on the Indian Prototype Fast Breeder Reactor (PFBR) being developed at Kalpakkam

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Captopril improves tumor nanomedicine delivery by increasing tumor blood perfusion and enlarging endothelial gaps in tumor blood vessels.

Science.gov (United States)

Zhang, Bo; Jiang, Ting; Tuo, Yanyan; Jin, Kai; Luo, Zimiao; Shi, Wei; Mei, Heng; Hu, Yu; Pang, Zhiqing; Jiang, Xinguo

2017-12-01

Poor tumor perfusion and unfavorable vessel permeability compromise nanomedicine drug delivery to tumors. Captopril dilates blood vessels, reducing blood pressure clinically and bradykinin, as the downstream signaling moiety of captopril, is capable of dilating blood vessels and effectively increasing vessel permeability. The hypothesis behind this study was that captopril can dilate tumor blood vessels, improving tumor perfusion and simultaneously enlarge the endothelial gaps of tumor vessels, therefore enhancing nanomedicine drug delivery for tumor therapy. Using the U87 tumor xenograft with abundant blood vessels as the tumor model, tumor perfusion experiments were carried out using laser Doppler imaging and lectin-labeling experiments. A single treatment of captopril at a dose of 100 mg/kg significantly increased the percentage of functional vessels in tumor tissues and improved tumor blood perfusion. Scanning electron microscopy of tumor vessels also indicated that the endothelial gaps of tumor vessels were enlarged after captopril treatment. Immunofluorescence-staining of tumor slices demonstrated that captopril significantly increased bradykinin expression, possibly explaining tumor perfusion improvements and endothelial gap enlargement. Additionally, imaging in vivo, imaging ex vivo and nanoparticle distribution in tumor slices indicated that after a single treatment with captopril, the accumulation of 115-nm nanoparticles in tumors had increased 2.81-fold with a more homogeneous distribution pattern in comparison to non-captopril treated controls. Finally, pharmacodynamics experiments demonstrated that captopril combined with paclitaxel-loaded nanoparticles resulted in the greatest tumor shrinkage and the most extensive necrosis in tumor tissues among all treatment groups. Taken together, the data from the present study suggest a novel strategy for improving tumor perfusion and enlarging blood vessel permeability simultaneously in order to improve

In vessel core melt progression phenomena

International Nuclear Information System (INIS)

Courtaud, M.

1993-01-01

For all light water reactor (LWR) accidents, including the so called severe accidents where core melt down can occur, it is necessary to determine the amount and characteristics of fission products released to the environment. For existing reactors this knowledge is used to evaluate the consequences and eventual emergency plans. But for future reactors safety authorities demand decrease risks and reactors designed in such a way that fission products are retained inside the containment, the last protective barrier. This requires improved understanding and knowledge of all accident sequences. In particular it is necessary to be able to describe the very complex phenomena occurring during in vessel core melt progression because they will determine the thermal and mechanical loads on the primary circuit and the timing of its rupture as well as the fission product source term. On the other hand, in case of vessel failure, knowledge of the physical and chemical state of the core melt will provide the initial conditions for analysis of ex-vessel core melt progression and phenomena threatening the containment. Finally a good understanding of in vessel phenomena will help to improve accident management procedures like Emergency Core Cooling System water injection, blowdown and flooding of the vessel well, with their possible adverse effects. Research and Development work on this subject was initiated a long time ago and is still in progress but now it must be intensified in order to meet the safety requirements of the next generation of reactors. Experiments, limited in scale, analysis of the TMI 2 accident which is a unique source of global information and engineering judgment are used to establish and assess physical models that can be implemented in computer codes for reactor accident analysis

Development and validation of a custom made indocyanine green fluorescence lymphatic vessel imager

Science.gov (United States)

Pallotta, Olivia J.; van Zanten, Malou; McEwen, Mark; Burrow, Lynne; Beesley, Jack; Piller, Neil

2015-06-01

Lymphoedema is a chronic progressive condition often producing significant morbidity. An in-depth understanding of an individual's lymphatic architecture is valuable both in the understanding of underlying pathology and for targeting and tailoring treatment. Severe lower limb injuries resulting in extensive loss of soft tissue require transposition of a flap consisting of muscle and/or soft tissue to close the defect. These patients are at risk of lymphoedema and little is known about lymphatic regeneration within the flap. Indocyanine green (ICG), a water-soluble dye, has proven useful for the imaging of lymphatic vessels. When injected into superficial tissues it binds to plasma proteins in lymph. By exposing the dye to specific wavelengths of light, ICG fluoresces with near-infrared light. Skin is relatively transparent to ICG fluorescence, enabling the visualization and characterization of superficial lymphatic vessels. An ICG fluorescence lymphatic vessel imager was manufactured to excite ICG and visualize real-time fluorescence as it travels through the lymphatic vessels. Animal studies showed successful ICG excitation and detection using this imager. Clinically, the imager has assisted researchers to visualize otherwise hidden superficial lymphatic pathways in patients postflap surgery. Preliminary results suggest superficial lymphatic vessels do not redevelop in muscle flaps.

Integrating Multiple Autonomous Underwater Vessels, Surface Vessels and Aircraft into Oceanographic Research Vessel Operations

Science.gov (United States)

McGillivary, P. A.; Borges de Sousa, J.; Martins, R.; Rajan, K.

2012-12-01

Autonomous platforms are increasingly used as components of Integrated Ocean Observing Systems and oceanographic research cruises. Systems deployed can include gliders or propeller-driven autonomous underwater vessels (AUVs), autonomous surface vessels (ASVs), and unmanned aircraft systems (UAS). Prior field campaigns have demonstrated successful communication, sensor data fusion and visualization for studies using gliders and AUVs. However, additional requirements exist for incorporating ASVs and UASs into ship operations. For these systems to be optimally integrated into research vessel data management and operational planning systems involves addressing three key issues: real-time field data availability, platform coordination, and data archiving for later analysis. A fleet of AUVs, ASVs and UAS deployed from a research vessel is best operated as a system integrated with the ship, provided communications among them can be sustained. For this purpose, Disruptive Tolerant Networking (DTN) software protocols for operation in communication-challenged environments help ensure reliable high-bandwidth communications. Additionally, system components need to have considerable onboard autonomy, namely adaptive sampling capabilities using their own onboard sensor data stream analysis. We discuss Oceanographic Decision Support System (ODSS) software currently used for situational awareness and planning onshore, and in the near future event detection and response will be coordinated among multiple vehicles. Results from recent field studies from oceanographic research vessels using AUVs, ASVs and UAS, including the Rapid Environmental Picture (REP-12) cruise, are presented describing methods and results for use of multi-vehicle communication and deliberative control networks, adaptive sampling with single and multiple platforms, issues relating to data management and archiving, and finally challenges that remain in addressing these technological issues. Significantly, the

Correlations of Nucleate Boiling Heat Transfer and Critical Heat Flux for External Reactor Vessel Cooling

International Nuclear Information System (INIS)

J. Yang; F. B. Cheung; J. L. Rempe; K. Y. Suh; S. B. Kim

2005-01-01

Four types of steady-state boiling experiments were conducted to investigate the efficacy of two distinctly different heat transfer enhancement methods for external reactor vessel cooling under severe accident conditions. One method involved the use of a thin vessel coating and the other involved the use of an enhanced insulation structure. By comparing the results obtained in the four types of experiments, the separate and integral effect of vessel coating and insulation structure were determined. Correlation equations were obtained for the nucleate boiling heat transfer and the critical heat flux. It was found that both enhancement methods were quite effective. Depending on the angular location, the local critical heat flux could be enhanced by 1.4 to 2.5 times using vessel coating alone whereas it could be enhanced by 1.8 to 3.0 times using an enhanced insulation structure alone. When both vessel coating and insulation structure were used simultaneously, the integral effect on the enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods

Testing of VVER reactor pressure vessels by TOFD method

International Nuclear Information System (INIS)

Skala, Z.; Vit, J.

2002-01-01

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

Superficial vessel reconstruction with a multiview camera system

Science.gov (United States)

Marreiros, Filipe M. M.; Rossitti, Sandro; Karlsson, Per M.; Wang, Chunliang; Gustafsson, Torbjörn; Carleberg, Per; Smedby, Örjan

2016-01-01

Abstract. We aim at reconstructing superficial vessels of the brain. Ultimately, they will serve to guide the deformation methods to compensate for the brain shift. A pipeline for three-dimensional (3-D) vessel reconstruction using three mono-complementary metal-oxide semiconductor cameras has been developed. Vessel centerlines are manually selected in the images. Using the properties of the Hessian matrix, the centerline points are assigned direction information. For correspondence matching, a combination of methods was used. The process starts with epipolar and spatial coherence constraints (geometrical constraints), followed by relaxation labeling and an iterative filtering where the 3-D points are compared to surfaces obtained using the thin-plate spline with decreasing relaxation parameter. Finally, the points are shifted to their local centroid position. Evaluation in virtual, phantom, and experimental images, including intraoperative data from patient experiments, shows that, with appropriate camera positions, the error estimates (root-mean square error and mean error) are ∼1  mm. PMID:26759814

Assessment of integrity for the pressure vessel internals of PWRs under blowdown loadings

International Nuclear Information System (INIS)

Geiss, M.; Benner, J.; Ludwig, A.

1984-01-01

In safety analysis of pressurized water reactors the loss-of-coolant accident plays a central role. Thereby a sudden break of a cold primary coolant pipe close to the reactor pressure vessel is postulated. The sudden pressure release of the primary system (blowdown) causes high dynamic loading on the pressure vessel internals. The resulting deformations must not impair shut down of the reactor and decay heat removal in an inadmissible way. For this assessment a blowdown analysis for a 1300 MW pressurized water reactor is carried out. These investigations are completed with a detailed stress analysis for the highly loaded core barrel clamping. The results show that the reactor pressure vessel internals are able to withstand blowdown loading. Even in case of a sudden and complete break of the primary coolant pipe the loading has to be twice as high to endanger the structural integrity. (orig.) [de

Effect of in-core instrumentation mounting location on external reactor vessel cooling

International Nuclear Information System (INIS)

Suh, Jungsoo; Ha, Huiun

2017-01-01

Highlights: • Numerical simulations were conducted for the evaluation of an IVR-ERVC application. • The ULPU-V experiment was simulated for the validation of numerical method. • The effect of ICI mounting location on an IVR-ERVC application was investigated. • TM-ICI is founded to be superior to BM-ICI for successful application of IVR-ERVC. - Abstract: The effect of in-core instrumentation (ICI) mounting location on the application of in-vessel corium retention through external reactor vessel cooling (IVR-ERVC), used to mitigate severe accidents in which the nuclear fuel inside the reactor vessel becomes molten, was investigated. Numerical simulations of the subcooled boiling flow within an advanced pressurized-water reactor (PWR) in IVR-ERVC applications were conducted for the cases of top-mounted ICI (TM-ICI) and bottom-mounted ICI (BM-ICI), using the commercially available computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the RPI model were used for turbulence closure and subcooled flow boiling, respectively. To validate the numerical method for IVR applications, numerical simulations of ULPU-V experiments were also conducted. The BM-ICI reactor vessel was modeled using a simplified design of an advanced PWR with BM-ICI; the TM-ICI counterpart was modeled by removing the ICI parts from the original geometry. It was found that TM-ICI was superior to BM-ICI for successful application of IVR-ERVC. For the BM-ICI case, the flow field was complicated because of the existence of ICIs and a significant temperature gradient was observed near the ICI nozzles on the lower part of the reactor vessel, where the ICIs were attached. These observations suggest that the existence of ICI below the reactor vessel hinders reactor vessel cooling.

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)

Comparative analysis of methods for extracting vessel network on breast MRI images

Science.gov (United States)

Gaizer, Bence T.; Vassiou, Katerina G.; Lavdas, Eleftherios; Arvanitis, Dimitrios L.; Fezoulidis, Ioannis V.; Glotsos, Dimitris T.

2017-11-01

Digital processing of MRI images aims to provide an automatized diagnostic evaluation of regular health screenings. Cancerous lesions are proven to cause an alteration in the vessel structure of the diseased organ. Currently there are several methods used for extraction of the vessel network in order to quantify its properties. In this work MRI images (Signa HDx 3.0T, GE Healthcare, courtesy of University Hospital of Larissa) of 30 female breasts were subjected to three different vessel extraction algorithms to determine the location of their vascular network. The first method is an experiment to build a graph over known points of the vessel network; the second algorithm aims to determine the direction and diameter of vessels at these points; the third approach is a seed growing algorithm, spreading selection to neighbors of the known vessel pixels. The possibilities shown by the different methods were analyzed, and quantitative measurements were performed. The data provided by these measurements showed no clear correlation with the presence or malignancy of tumors, based on the radiological diagnosis of skilled physicians.

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.

Magnetic resonance imaging of water ascent in embolized xylem vessels of grapevine stem segments

Science.gov (United States)

Mingtao Wang; Melvin T. Tyree; Roderick E. Wasylishen

2013-01-01

Temporal and spatial information about water refilling of embolized xylem vessels and the rate of water ascent in these vessels is critical for understanding embolism repair in intact living vascular plants. High-resolution 1H magnetic resonance imaging (MRI) experiments have been performed on embolized grapevine stem segments while they were...

Simulation of In-Vessel Corium Retention through External Reactor Vessel Cooling for SMART using SIMPLE

Energy Technology Data Exchange (ETDEWEB)

Jang, Jin-Sung; Son, Donggun; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Thermal load analysis from the corium pool to the outer reactor vessel in the lower plenum of the reactor vessel is necessary to evaluate the effect of the IVR-ERVC during a severe accident for SMART. A computational code called SIMPLE (Sever Invessel Melt Progression in Lower plenum Environment) has been developed for analyze transient behavior of molten corium in the lower plenum, interaction between corium and coolant, and heat-up and ablation of reactor vessel wall. In this study, heat load analysis of the reactor vessel for SMART has been conducted using the SIMPLE. Transient behavior of the molten corium in the lower plenum and IVR-ERVC for SMART has been simulated using SIMPLE. Heat flux from the corium pool to the outer reactor vessel is concentrated in metallic layer by the focusing effect. As a result, metallic layer shows higher temperature than the oxidic layer. Also, vessel wall of metallic layer has been ablated by the high in-vessel temperature. Ex-vessel temperature of the metallic layer was maintained 390 K and vessel thickness was maintained 14 cm. It means that the reactor vessel integrity is maintained by the IVR-ERVC.

A computational algorithm addressing how vessel length might depend on vessel diameter

Science.gov (United States)

Jing Cai; Shuoxin Zhang; Melvin T. Tyree

2010-01-01

The objective of this method paper was to examine a computational algorithm that may reveal how vessel length might depend on vessel diameter within any given stem or species. The computational method requires the assumption that vessels remain approximately constant in diameter over their entire length. When this method is applied to three species or hybrids in the...

Experiments and Analysis of Close-Shot Identification of On-Branch Citrus Fruit with RealSense

Directory of Open Access Journals (Sweden)

Jizhan Liu

2018-05-01

Full Text Available Fruit recognition based on depth information has been a hot topic due to its advantages. However, the present equipment and methods cannot meet the requirements of rapid and reliable recognition and location of fruits in close shot for robot harvesting. To solve this problem, we propose a recognition algorithm for citrus fruit based on RealSense. This method effectively utilizes depth-point cloud data in a close-shot range of 160 mm and different geometric features of the fruit and leaf to recognize fruits with a intersection curve cut by the depth-sphere. Experiments with close-shot recognition of six varieties of fruit under different conditions were carried out. The detection rates of little occlusion and adhesion were from 80–100%. However, severe occlusion and adhesion still have a great influence on the overall success rate of on-branch fruits recognition, the rate being 63.8%. The size of the fruit has a more noticeable impact on the success rate of detection. Moreover, due to close-shot near-infrared detection, there was no obvious difference in recognition between bright and dark conditions. The advantages of close-shot limited target detection with RealSense, fast foreground and background removal and the simplicity of the algorithm with high precision may contribute to high real-time vision-servo operations of harvesting robots.

Acoustic isolation vessel for measurement of the background noise in microphones

Science.gov (United States)

Ngo, Kim C. T.; Zuckerwar, Allan J.

1993-01-01

An acoustic isolation vessel has been developed to measure the background noise in microphones. The test microphone is installed in an inner vessel, which is suspended within an outer vessel, and the intervening air space is evacuated to a high vacuum. An analytical expression for the transmission coefficient is derived, based on a five-media model, and compared to experiment. At an isolation vacuum of 5 x 10 exp -6 Torr the experimental transmission coefficient was found to be lower than -155 dB at frequencies ranging from 40 to 1200 Hz. Measurements of the A-weighted noise levels of commercial condenser microphones of four different sizes show good agreement with published values.

Investigation of flow stabilization in a compact reactor vessel of a FBR. Flow visualization in a reactor vessel

International Nuclear Information System (INIS)

Sato, Hiroyuki; Igarashi, Minoru; Kimura, Nobuyuki; Kamide, Hideki

2002-01-01

In the feasibility studies of Commercialized Fast Breeder Reactor Cycle System, a compact reactor vessel is considered from economical improvement point of a sodium cooled loop type fast reactor. The flow field was visualized by water experiment for a reactor vessel with 'a column type UIS (Upper Internal Structure)', which has a slit for fuel handling mechanism and is useful for a compact fast reactor. In this research, the 1/20 scale test equipment using water was made to understand coolant flow through a slit of a column type UIS' and fundamental behavior of reactor upper plenum flow. In the flow visualization tests, tracer particles were added in the water, and illuminated by the slit-shaped pulse laser. The flow visualization image was taken with a CCD camera. We obtained fluid velocity vectors from the visualization image using the Particle Imaging Velocimetry (PIV). The results are as follows. 1. Most of coolant flow through a slit of 'column type UIS' arrived the dip plate directly. In the opposite side of a slit, most of coolant flowed toward reactor vessel wall before it arrived the dip plate. 2. The PIV was useful to measure the flow field in the reactor vessel. The obtained velocity field was consistent with the flow visualization result. 3. The jet through the UIS slit was dependent on the UIS geometry. There is a possibility to control the jet by the UIS geometry. (author)

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.

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.)

Cruise ship environmental hygiene and the risk of norovirus infection outbreaks: an objective assessment of 56 vessels over 3 years.

Science.gov (United States)

Carling, Philip C; Bruno-Murtha, Lou Ann; Griffiths, Jeffrey K

2009-11-01

Norovirus infection outbreaks (NoVOs) occur frequently in closed populations, such as cruise ship passengers. Environmental contamination is believed to play an important role in NoVO propagation. Trained health care professionals covertly evaluated the thoroughness of disinfection cleaning (TDC) of 6 standardized objects (toilet seat, flush handle or button, toilet stall inner handhold, stall inner door handle, restroom inner door handle, and baby changing table surfaces) with high potential for fecal contamination in cruise ship public restrooms, by means of a previously validated novel targeting method. Fifty-six cruise ships (approximately 30% of 180 vessels operated by 9 large cruise lines) were evaluated from July 2005 through August 2008. Overall, 37% (range, 4%-100%; 95% confidence interval, 29.2%-45.4%) of 8344 objects in 273 randomly selected public restrooms were cleaned daily. The TDC did not differ by cruise line and did not correlate with the Centers for Disease Control and Prevention (CDC) Vessel Sanitation Program inspection scores (r(2), .002; P = .75). More than half the vessels had overall TDC scores ships had near-perfect CDC sanitation scores. The mean TDC of the 3 ships evaluated within 4 months before a NoVO (10.3%) was substantially less than the mean TDC of the 40 ships that did not experience NoVOs (40.4%) (P ships found that only 37% of selected toilet area objects were cleaned on a daily basis. Low TDC scores may predict subsequent NoVO-prone vessels. Enhanced public restroom cleaning may prevent or moderate NoVOs on cruise ships.

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

Variable flow controls of closed system pumps for energy savings in maritime power systems

DEFF Research Database (Denmark)

Su, Chun-Lien; Liao, Chi-Hsiang; Chou, Tso-Chu

2016-01-01

and field tests of a practical auxiliary boiler feed water management system on a commercial vessel. It is proved that the proposed method can maintain constant water pressure for closed system pumps and provide an efficient way to measure energy savings and maintenance benefits. The results serve......Pumps are extensively used in maritime industries as marine vessels utilize a wide range of pumps and pumping techniques to transfer and distribute all types of air and fluids. The electrical energy consumed by the various motors accounts for about 70% of a vessel’s total power consumption......, and this presents a problem in unique marine environments. Such situations are especially conducive to energy-saving strategies using variable frequency drives (VFDs) in centrifugal load service. This paper presents the design and results of applying variable frequency constant pressure technology in closed system...

Requirements for thermal insulation on prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Neylan, A.J.; Wistrom, J.D.

1979-01-01

During the past decade, extensive design, construction, and operating experience on concrete pressure vessels for gas-cooled reactor applications has accumulated. Excellent experience has been obtained to date on the structural components (concrete, prestressing systems, liners, penetrations, and closures) and the thermal insulation. Three fundamentally different types of insulation systems have been employed to ensure the satisfactory performance of this component, which is critical to the overall success of the prestressed concrete reactor vessel (PCRV). Although general design criteria have been published, the requirements for design, materials, and construction are not rigorously addressed in any national or international code. With the more onerous design conditions being imposed by advanced reactor systems, much greater attention has been directed to advance the state of the art of insulation systems for PCRVs. This paper addresses some of the more recent developments in this field being performed by General Atomic Company and others. (author)

Defining an adequate sample of earlywood vessels for retrospective injury detection in diffuse-porous species.

Directory of Open Access Journals (Sweden)

Estelle Arbellay

Full Text Available Vessels of broad-leaved trees have been analyzed to study how trees deal with various environmental factors. Cambial injury, in particular, has been reported to induce the formation of narrower conduits. Yet, little or no effort has been devoted to the elaboration of vessel sampling strategies for retrospective injury detection based on vessel lumen size reduction. To fill this methodological gap, four wounded individuals each of grey alder (Alnus incana (L. Moench and downy birch (Betula pubescens Ehrh. were harvested in an avalanche path. Earlywood vessel lumina were measured and compared for each tree between the injury ring built during the growing season following wounding and the control ring laid down the previous year. Measurements were performed along a 10 mm wide radial strip, located directly next to the injury. Specifically, this study aimed at (i investigating the intra-annual duration and local extension of vessel narrowing close to the wound margin and (ii identifying an adequate sample of earlywood vessels (number and intra-ring location of cells attesting to cambial injury. Based on the results of this study, we recommend analyzing at least 30 vessels in each ring. Within the 10 mm wide segment of the injury ring, wound-induced reduction in vessel lumen size did not fade with increasing radial and tangential distances, but we nevertheless advise favoring early earlywood vessels located closest to the injury. These findings, derived from two species widespread across subarctic, mountainous, and temperate regions, will assist retrospective injury detection in Alnus, Betula, and other diffuse-porous species as well as future related research on hydraulic implications after wounding.

Albatrosses following fishing vessels: how badly hooked are they on an easy meal?

Directory of Open Access Journals (Sweden)

José P Granadeiro

Full Text Available Fisheries have major impacts on seabirds, both by changing food availability and by causing direct mortality of birds during trawling and longline setting. However, little is known about the nature and the spatial-temporal extent of the interactions between individual birds and vessels. By studying a system in which we had fine-scale data on bird movements and activity, and near real-time information on vessel distribution, we provide new insights on the association of a threatened albatross with fisheries. During early chick-rearing, black-browed albatrosses Thalassarche melanophris from two different colonies (separated by only 75 km showed significant differences in the degree of association with fisheries, despite being nearly equidistant to the Falklands fishing fleet. Most foraging trips from either colony did not bring tracked individuals close to vessels, and proportionally little time and foraging effort was spent near ships. Nevertheless, a few individuals repeatedly visited fishing vessels, which may indicate they specialise on fisheries-linked food sources and so are potentially more vulnerable to bycatch. The evidence suggests that this population has little reliance on fisheries discards at a critical stage of its nesting cycle, and hence measures to limit fisheries waste on the Patagonian shelf that also reduce vessel attractiveness and the risk of incidental mortality, would be of high overall conservation benefit.

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

Closed-loop control of renal perfusion pressure in physiological experiments.

Science.gov (United States)

Campos-Delgado, D U; Bonilla, I; Rodríguez-Martínez, M; Sánchez-Briones, M E; Ruiz-Hernández, E

2013-07-01

This paper presents the design, experimental modeling, and control of a pump-driven renal perfusion pressure (RPP)-regulatory system to implement precise and relatively fast RPP regulation in rats. The mechatronic system is a simple, low-cost, and reliable device to automate the RPP regulation process based on flow-mediated occlusion. Hence, the regulated signal is the RPP measured in the left femoral artery of the rat, and the manipulated variable is the voltage applied to a dc motor that controls the occlusion of the aorta. The control system is implemented in a PC through the LabView software, and a data acquisition board NI USB-6210. A simple first-order linear system is proposed to approximate the dynamics in the experiment. The parameters of the model are chosen to minimize the error between the predicted and experimental output averaged from eight input/output datasets at different RPP operating conditions. A closed-loop servocontrol system based on a pole-placement PD controller plus dead-zone compensation was proposed for this purpose. First, the feedback structure was validated in simulation by considering parameter uncertainty, and constant and time-varying references. Several experimental tests were also conducted to validate in real time the closed-loop performance for stepwise and fast switching references, and the results show the effectiveness of the proposed automatic system to regulate the RPP in the rat, in a precise, accurate (mean error less than 2 mmHg) and relatively fast mode (10-15 s of response time).

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)

ALICE HMPID Radiator Vessel

CERN Document Server

2003-01-01

View of the radiator vessels of the ALICE/HMPID mounted on the support frame. Each HMPID module is equipped with 3 indipendent radiator vessels made out of neoceram and fused silica (quartz) windows glued together. The spacers inside the vessel are needed to stand the hydrostatic pressure. http://alice-hmpid.web.cern.ch/alice-hmpid

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.

Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.

1988-05-01

The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of 1/6. Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates.

Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Zahoor, A.

1988-01-01

The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of 1/6. Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates. (orig.)

Conducting Closed Habitation Experiments: Experience from the Lunar Mars Life Support Test Project

Science.gov (United States)

Barta, Daniel J.; Edeen, Marybeth A.; Henninger, Donald L.

2006-01-01

The Lunar-Mars Life Support Test Project (LMLSTP) was conducted from 1995 through 1997 at the National Aeronautics and Space Administration s (NASA) Johnson Space Center (JSC) to demonstrate increasingly longer duration operation of integrated, closed-loop life support systems that employed biological and physicochemical techniques for water recycling, waste processing, air revitalization, thermal control, and food production. An analog environment for long-duration human space travel, the conditions of isolation and confinement also enabled studies of human factors, medical sciences (both physiology and psychology) and crew training. Four tests were conducted, Phases I, II, IIa and III, with durations of 15, 30, 60 and 91 days, respectively. The first phase focused on biological air regeneration, using wheat to generate enough oxygen for one experimental subject. The systems demonstrated in the later phases were increasingly complex and interdependent, and provided life support for four crew members. The tests were conducted using two human-rated, atmospherically-closed test chambers, the Variable Pressure Growth Chamber (VPGC) and the Integrated Life Support Systems Test Facility (ILSSTF). Systems included test articles (the life support hardware under evaluation), human accommodations (living quarters, kitchen, exercise equipment, etc.) and facility systems (emergency matrix system, power, cooling, etc.). The test team was managed by a lead engineer and a test director, and included test article engineers responsible for specific systems, subsystems or test articles, test conductors, facility engineers, chamber operators and engineering technicians, medical and safety officers, and science experimenters. A crew selection committee, comprised of psychologists, engineers and managers involved in the test, evaluated male and female volunteers who applied to be test subjects. Selection was based on the skills mix anticipated for each particular test, and utilized

Reactor vessel sealing plug

International Nuclear Information System (INIS)

Dooley, R.A.

1986-01-01

This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

Vessel Eddy Current Measurement for the National Spherical Torus Experiment (NSTX)

International Nuclear Information System (INIS)

Gates, D.A.; Menard, J.; Marsala, R.

2004-01-01

A simple analog circuit that measures the NSTX axisymmetric eddy current distribution has been designed and constructed. It is based on simple circuit model of the NSTX vacuum vessel that was calibrated using a special axisymmetric eddy current code which was written so that accuracy was maintained in the vicinity of the current filaments. The measurement and the model have been benchmarked against data from numerous vacuum shots and they are in excellent agreement. This is an important measurement that helps give more accurate equilibrium reconstructions

Dynamics of micro-bubble sonication inside a phantom vessel

KAUST Repository

Qamar, Adnan; Samtaney, Ravi; Bull, Joseph L.

2013-01-01

A model for sonicated micro-bubble oscillations inside a phantom vessel is proposed. The model is not a variant of conventional Rayleigh-Plesset equation and is obtained from reduced Navier-Stokes equations. The model relates the micro-bubble oscillation dynamics with geometric and acoustic parameters in a consistent manner. It predicts micro-bubble oscillation dynamics as well as micro-bubble fragmentation when compared to the experimental data. For large micro-bubble radius to vessel diameter ratios, predictions are damped, suggesting breakdown of inherent modeling assumptions for these cases. Micro-bubble response with acoustic parameters is consistent with experiments and provides physical insight to the micro-bubble oscillation dynamics.

Image processing algorithm for robot tracking in reactor vessel

International Nuclear Information System (INIS)

Kim, Tae Won; Choi, Young Soo; Lee, Sung Uk; Jeong, Kyung Min; Kim, Nam Kyun

2011-01-01

In this paper, we proposed an image processing algorithm to find the position of an underwater robot in the reactor vessel. Proposed algorithm is composed of Modified SURF(Speeded Up Robust Feature) based on Mean-Shift and CAMSHIFT(Continuously Adaptive Mean Shift Algorithm) based on color tracking algorithm. Noise filtering using luminosity blend method and color clipping are preprocessed. Initial tracking area for the CAMSHIFT is determined by using modified SURF. And then extracting the contour and corner points in the area of target tracked by CAMSHIFT method. Experiments are performed at the reactor vessel mockup and verified to use in the control of robot by visual tracking

Dynamics of micro-bubble sonication inside a phantom vessel

KAUST Repository

Qamar, Adnan

2013-01-10

A model for sonicated micro-bubble oscillations inside a phantom vessel is proposed. The model is not a variant of conventional Rayleigh-Plesset equation and is obtained from reduced Navier-Stokes equations. The model relates the micro-bubble oscillation dynamics with geometric and acoustic parameters in a consistent manner. It predicts micro-bubble oscillation dynamics as well as micro-bubble fragmentation when compared to the experimental data. For large micro-bubble radius to vessel diameter ratios, predictions are damped, suggesting breakdown of inherent modeling assumptions for these cases. Micro-bubble response with acoustic parameters is consistent with experiments and provides physical insight to the micro-bubble oscillation dynamics.

The response of pressure vessel steel specimens on drop weight loading

International Nuclear Information System (INIS)

Winkler, S.; Kalthoff, J.F.; Gerscha, A.

1979-01-01

Load records obtained in instrumented impact tests in general are disturbed by inertia effects. The influence of mechanical damping provisions on these disturbing inertia effects is investigated. Precracked bend specimens are dynamically loaded in a drop weight testing system. The specimens of size 620 mm x 150 mm (25 mm or 50 mm thick) were machined from the pressure vessel steel 22 NiMoCr 37 which was heat treated to achieve a specially hardened condition. The tests were performed at two different low temperatures. The impact velocity was about 4 m/s. As it is usual in instrumented impact testing, the load at the tup of the impining striker is recorded as a function of time during the impact process. In addition the specimen is instrumented by a strain gage close to the crack tip in order to directly measure the stress intensification. Experiments were performed under pure and damped impact conditions. Damping was achieved by utilizing a soft aluminum plate between the striker and the specimen. (orig.)

Parametrically defined cerebral blood vessels as non-invasive blood input functions for brain PET studies

International Nuclear Information System (INIS)

Asselin, Marie-Claude; Cunningham, Vincent J; Amano, Shigeko; Gunn, Roger N; Nahmias, Claude

2004-01-01

A non-invasive alternative to arterial blood sampling for the generation of a blood input function for brain positron emission tomography (PET) studies is presented. The method aims to extract the dimensions of the blood vessel directly from PET images and to simultaneously correct the radioactivity concentration for partial volume and spillover. This involves simulation of the tomographic imaging process to generate images of different blood vessel and background geometries and selecting the one that best fits, in a least-squares sense, the acquired PET image. A phantom experiment was conducted to validate the method which was then applied to eight subjects injected with 6-[ 18 F]fluoro-L-DOPA and one subject injected with [ 11 C]CO-labelled red blood cells. In the phantom study, the diameter of syringes filled with an 11 C solution and inserted into a water-filled cylinder were estimated with an accuracy of half a pixel (1 mm). The radioactivity concentration was recovered to 100 ± 4% in the 8.7 mm diameter syringe, the one that most closely approximated the superior sagittal sinus. In the human studies, the method systematically overestimated the calibre of the superior sagittal sinus by 2-3 mm compared to measurements made in magnetic resonance venograms on the same subjects. Sources of discrepancies related to the anatomy of the blood vessel were found not to be fundamental limitations to the applicability of the method to human subjects. This method has the potential to provide accurate quantification of blood radioactivity concentration from PET images without the need for blood samples, corrections for delay and dispersion, co-registered anatomical images, or manually defined regions of interest

[Key vessels assessment and operation highlights in laparoscopic extended right hemicolectomy].

Science.gov (United States)

Wang, Hao; Zhao, Quanquan

2018-03-25

Laparoscopic radical colectomies have been more widely used gradually, among which laparoscopic extended right hemicolectomy is considered as the most difficult procedure. The difficulty of extended right hemicolectomy lies in the need to dissect lymph nodes along the superior mesenteric vein (SMV) and disconnect numerous and possible aberrant vessels. To address this problem, we emphasize two points in key vessel assessment: getting familiar with the anatomy along the medial-to-lateral approach and having a good understanding about the preoperative imaging presentations. An accurately preoperative imaging assessment by abdominal enhanced CT can help the surgeon understand the relative position of the key vessels to be dealt with during operation and the situation of the possible aberrant vessels so as to guide the procedure more effectively and facilitate the prevention and management of the intraoperative complications. During operation, the operator should pay special attention to the management of the vessels in the ileocolic vessel region, Henle's trunk and middle colon vessels. The operation highlights of the key vessels are as follows: (1) The ileocolic vessels: identifying the Toldt's gap correctly and opening the vascular sheath of the SMV securely; making sure that the duodenum is well protected. (2) Henle's trunk: dissecting along the surface of the Henle's trunk; preserving the anterior superior pancreaticoduodenal vein (ASPDV) and main trunk of the Henle's trunk; disconnecting the roots of the right colic vein (RCV) and right gastroepiploic vein (RGEV), and then dissecting lymph nodes along the surface of the pancreas. (3) The middle colon vessels: identifying the root of the middle colon vessel along the lower edge of the pancreas; avoiding entering behind the pancreas; mobilizing the transverse mesocolon sufficiently along the surface of the pancreas. Finally, we discuss and analyze the disputes currently existing in laparoscopic extended right

Synergistic effect of fly ash in in-vessel composting of biomass and kitchen waste.

Science.gov (United States)

Manyapu, Vivek; Mandpe, Ashootosh; Kumar, Sunil

2018-03-01

The present study aims to utilize coal fly ash for its property to adsorb heavy metals and thus reducing the bioavailability of the metals for plant uptake. Fly ash was incorporated into the in-vessel composting system along with organic waste. The in-vessel composting experiments were conducted in ten plastic vessels of 15 L capacity comprising varying proportions of biomass waste, kitchen waste and fly ash. In this study, maximum degradation of organic matter was observed in Vessel 3 having k value of 0.550 d -1 . In vessel 10, 20% fly ash with a combination of 50% biomass waste and 30% kitchen waste along with the addition of 5% jaggery as an additive produced the best outcome with least organic matter (%C) loss and lowest value of rate constant (k). Copyright © 2017 Elsevier Ltd. All rights reserved.

Cooling of particulate debris beds: analysis of the initial D-series experiments

International Nuclear Information System (INIS)

Rivard, J.B.

1978-01-01

In an effort to provide basic data on the cooling of fast reactor debris, three in-pile experiments employing oxide fuel particulate in liquid sodium were completed in late 1977. Preliminary results from these experiments were reported shortly after their completion at the Third Post-Accident Heat Removal Information Exchange, at Argonne National Laboratory. In these experiments, a distribution of 100 μm to 1000 μm-sized particles of enriched UO 2 was fission-heated to simulate decay-heated debris. In each experiment, the UO 2 particles were contained in a closed, flat-bottomed vessel 012 mm in diameter which was insulated on the diameter and bottom. Sufficient sodium was included to saturate the bed of particles and to provide a volume of bulk sodium above the bed at a controlled temperature. Parameters of interest in the experiments are given

[Pathophysiological changes of umbilical vessels in intrauterine growth restriction].

Science.gov (United States)

Jakó, Mária; Surányi, Andrea; Kaiser, László; Domokos, Dóra; Gáspár, Róbert; Bártfai, György

2014-12-14

The prevalence of intrauterine growth restriction is 4-5000/100,000 births, and they give the majority of perinatal morbidity. The aim of the authors was to compare the pathomorphologic data and vasoreactivity of umbilical vessels and placenta of small for date newborns to that of the normal pregnancies. Samples of the umbilical cord and placenta were divided into case and control groups. Two 10 cm long segments were cut of the umbilical cord at placental insertion. Tissue bath experiment was performed on umbilical vessels and pathomorphologic data were collected according to the Royal College of Pathologists' protocol. After the development of basal tone, oxytocin and desmopressin did not enhance the vascular contraction, but the pathomorphological and ultrasonographic data were significantly different in the two groups. The results indicate that umbilical vessels might not have oxytocin or vasopressin receptors. The pathomorphologic and flowmetric differences could be the causes of small birth weight.

Transferability of results of PTS experiments to the integrity assessment of reactor pressure vessels

International Nuclear Information System (INIS)

Roos, E.; Eisele, U.; Stumpfrock, L.

1997-01-01

The integrity assessment of the reactor pressure vessel (RPV) is based on the fracture mechanics concept as provided in the code. However this concept covers only the linear-elastic fracture mechanics regime on the basis of the reference temperature RT NDT as derived from charpy impact and drop-weight test. The conservatism of this concept was demonstrated for a variety of different materials covering optimized and lower bound material states with regard to unirradiated and irradiated conditions. For the elastic-plastic regime, methodologies have been developed to describe ductile crack initiation and stable crack growth. The transferability of both, the linear-elastic and elastic-plastic fracture mechanics concept was investigated with the help of large scale specimens focusing on complex loading situations as they result from postulated thermal shock events for the RPV. A series of pressurized thermal shock (PTS) experiments were performed in which the applicability of the fracture mechanics parameters derived from small scale specimen testing could be demonstrated. This includes brittle (static and dynamic) crack initiation and crack arrest in the low charpy energy regime as well as stable crack initiation, stable crack growth and crack arrest in the upper shelf toughness regime. The paper provides the basic material data, the load paths, representative for large complex components as well as experimental and theoretical results of PTS experiments. From these data it can be concluded that the available fracture mechanics concepts can be used to describe the component behavior under transient loading conditions. (author). 26 refs, 12 figs, 1 tab

Transferability of results of PTS experiments to the integrity assessment of reactor pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Roos, E; Eisele, U; Stumpfrock, L [MPA Stuttgart (Germany)

1997-09-01

The integrity assessment of the reactor pressure vessel (RPV) is based on the fracture mechanics concept as provided in the code. However this concept covers only the linear-elastic fracture mechanics regime on the basis of the reference temperature RT{sub NDT} as derived from charpy impact and drop-weight test. The conservatism of this concept was demonstrated for a variety of different materials covering optimized and lower bound material states with regard to unirradiated and irradiated conditions. For the elastic-plastic regime, methodologies have been developed to describe ductile crack initiation and stable crack growth. The transferability of both, the linear-elastic and elastic-plastic fracture mechanics concept was investigated with the help of large scale specimens focusing on complex loading situations as they result from postulated thermal shock events for the RPV. A series of pressurized thermal shock (PTS) experiments were performed in which the applicability of the fracture mechanics parameters derived from small scale specimen testing could be demonstrated. This includes brittle (static and dynamic) crack initiation and crack arrest in the low charpy energy regime as well as stable crack initiation, stable crack growth and crack arrest in the upper shelf toughness regime. The paper provides the basic material data, the load paths, representative for large complex components as well as experimental and theoretical results of PTS experiments. From these data it can be concluded that the available fracture mechanics concepts can be used to describe the component behavior under transient loading conditions. (author). 26 refs, 12 figs, 1 tab.

Containment vessel drain system

Science.gov (United States)

Harris, Scott G.

2018-01-30

A system for draining a containment vessel may include a drain inlet located in a lower portion of the containment vessel. The containment vessel may be at least partially filled with a liquid, and the drain inlet may be located below a surface of the liquid. The system may further comprise an inlet located in an upper portion of the containment vessel. The inlet may be configured to insert pressurized gas into the containment vessel to form a pressurized region above the surface of the liquid, and the pressurized region may operate to apply a surface pressure that forces the liquid into the drain inlet. Additionally, a fluid separation device may be operatively connected to the drain inlet. The fluid separation device may be configured to separate the liquid from the pressurized gas that enters the drain inlet after the surface of the liquid falls below the drain inlet.

Assessing the Efficiency of Small-Scale and Bottom Trawler Vessels in Greece

Directory of Open Access Journals (Sweden)

Dario Pinello

2016-07-01

Full Text Available This study explores the technical and scale efficiency of two types of Greek fishing vessels, small-scale vessels and bottom trawlers, using a bias-corrected input-oriented Data Envelopment Analysis model. Moreover, the associations between efficiency scores and vessel’s and skipper’s characteristics are also explored. The results indicate that small-scale vessels achieve a very low average technical efficiency score (0.42 but a much higher scale efficiency score (0.81. Conversely, bottom trawlers achieve lower scale but higher technical efficiency scores (0.68 and 0.73, respectively. One important finding of this study is that the technical efficiency of small-scale vessels, in contrast to trawlers, is positively associated with the experience of the skipper. In a looser context, it can be said that small-scale fisheries mainly rely on skill, whereas bottom trawlers rely more on technology. This study concludes that there is space for improvement in efficiency, mainly for small-scale vessels, which could allow the achievement of the same level of output by using reduced inputs.

The POST trial: initial post-market experience of the Penumbra system: revascularization of large vessel occlusion in acute ischemic stroke in the United States and Europe.

Science.gov (United States)

Tarr, Robert; Hsu, Dan; Kulcsar, Zsolt; Bonvin, Christophe; Rufenacht, Daniel; Alfke, Karsten; Stingele, Robert; Jansen, Olav; Frei, Donald; Bellon, Richard; Madison, Michael; Struffert, Tobias; Dorfler, Arnd; Grunwald, Iris Q; Reith, Wolfgang; Haass, Anton

2010-12-01

The purpose of this study was to assess the initial post-market experience of the device and how it is compared with the Penumbra Pivotal trial used to support the 510k application. A retrospective case review of 157 consecutive patients treated with the Penumbra system at seven international centers was performed. Primary endpoints were revascularization of the target vessel (TIMI score of 2 or 3), good functional outcome as defined by a modified Rankin scale (mRS) score of ≤2 and incidence of procedural serious adverse events. Results were compared with those of the Penumbra pivotal trial. A total of 157 vessels were treated. Mean baseline values at enrollment were: age 65 years, NIHSS score 16. After use of the Penumbra system, 87% of the treated vessels were revascularized to TIMI 2 (54%) or 3 (33%) as compared with 82% reported in the Pivotal trial. Nine procedural serious adverse events were reported in 157 patients (5.7%). All-cause mortality was 20% (32/157), and 41% had a mRS of ≤2 at 90-day follow-up as compared with only 25% in the Pivotal trial. Patients who were successfully revascularized by the Penumbra system had significantly better outcomes than those who were not. Initial post-market experience of the Penumbra system revealed that the revascularization rate and safety profile of the device are comparable to those reported in the Pivotal trial. However, the proportion of patients who had good functional outcome was higher than expected.

Fatigue and fracture mechanics in pressure vessels and piping. PVP-Volume 304

International Nuclear Information System (INIS)

Mehta, H.S.; Wilkowski, G.; Takezono, S.; Bloom, J.; Yoon, K.; Aoki, S.; Rahman, S.; Nakamura, T.; Brust, F.; Yoshimura, S.

1995-01-01

Fracture mechanics and fatigue evaluations are an important part of the structural integrity analyses to assure safe operation of pressure vessels and piping components during their service life. The paper presented in this volume illustrate the application of fatigue and fracture mechanics techniques to assess the structural integrity of a wide variety of Pressure Vessels and Piping components. The papers are organized in six sections: (1) fatigue and fracture--vessels; (2) fatigue and fracture--piping; (3) fatigue and fracture--material property evaluations; (4) constraint effects in fracture mechanics; (5) probabilistic fracture mechanics analyses; and (6) user's experience with failure assessment diagrams. Separate abstracts were prepared for most of the papers in this book

Belgian experience in applying the {open_quotes}leak-before-break{close_quotes} concept to the primary loop piping

Energy Technology Data Exchange (ETDEWEB)

Gerard, R.; Malekian, C.; Meessen, O. [Tractebel Energy Engineering, Brussels (Belgium)

1997-04-01

The Leak Before Break (LBB) concept allows to eliminate from the design basis the double-ended guillotine break of the primary loop piping, provided it can be demonstrated by a fracture mechanics analysis that a through-wall flaw, of a size giving rise to a leakage still well detectable by the plant leak detection systems, remains stable even under accident conditions (including the Safe Shutdown Earthquake (SSE)). This concept was successfully applied to the primary loop piping of several Belgian Pressurized Water Reactor (PWR) units, operated by the Utility Electrabel. One of the main benefits is to permit justification of supports in the primary loop and justification of the integrity of the reactor pressure vessel and internals in case of a Loss Of Coolant Accident (LOCA) in stretch-out conditions. For two of the Belgian PWR units, the LBB approach also made it possible to reduce the number of large hydraulic snubbers installed on the primary coolant pumps. Last but not least, the LBB concept also facilitates the steam generator replacement operations, by eliminating the need for some pipe whip restraints located close to the steam generator. In addition to the U.S. regulatory requirements, the Belgian safety authorities impose additional requirements which are described in details in a separate paper. An novel aspect of the studies performed in Belgium is the way in which residual loads in the primary loop are taken into account. Such loads may result from displacements imposed to close the primary loop in a steam generator replacement operation, especially when it is performed using the {open_quote}two cuts{close_quotes} technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented.

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)

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

Overview of Computational Fluid Dynamics (CFD) simulation of stirred vessel

International Nuclear Information System (INIS)

Mohd Rizal Mamat; Azraf Azman; Anwar Abdul Rahman; Noraishah Othman

2010-01-01

Stirred vessel is one of many widely used equipment in industrial process and chemical industry. The design of stirred vessel typically follows a certain standard chemical engineering practice that may also involve empirical data acquired from experiments. However the design may still take a different route which is computational engineering simulation and analysis. CFD has been identified as one of the possible tools for such purposes. CFD enables the flow fields variables such as velocity, temperature and pressure in the whole computational domain to be obtained and as such it presents an advantage over the experimental setup. (author)

Prediction and experimental verification of performance of box type solar cooker. Part II: Cooking vessel with depressed lid

International Nuclear Information System (INIS)

Reddy, Avala Raji; Rao, A.V. Narasimha

2008-01-01

Our previous article (Part I) discussed the theoretical and experimental study of the performance boost obtained by a cooking vessel with central cylindrical cavity on lugs when compared to that of a conventional cylindrical vessel on floor/lugs. This article compares the performance of the cooking vessel with depressed lid on lugs with that of the conventional vessel on lugs. A mathematical model is presented to understand the heat flow process to the cooking vessel and, thereby, to the food material. It is found from the experiments that the cooking vessel with depressed lid results in higher temperature of the thermic fluid loaded in the cooking vessel compared to that of the thermic fluid kept in the conventional vessel when both are placed on lugs. Similar results were obtained by modeling the process mathematically. The average improvement of performance of the vessel with depressed lid is found to be 8.4% better than the conventional cylindrical vessel

Development of a Compact and Efficient Ice Thermal Energy Storage Vessel

Science.gov (United States)

Sasaguchi, Kengo; Ishikawa, Masatoshi; Muta, Kenji; Yoshino, Kiyotaka; Hayashi, Hiroko; Baba, Yoshiyuki

In the present study, the authors propose the use of a low concentration aqueous solution as phase change material for static-type ice-storage-vessels, instead of pure water commonly used today. If an aqueous solution with low concentration is used, even when a large amount of solution (aqueous ethylene glycol in this study) is solidified and bridging of ice developed around cold tubes occurs, the pressure increase could be prevented by the existence of a continuous liquid phase in the solid-liquid two-phase layer (mushy layer) which opens to an air gap at the top of a vessel. Therefore, one can continue to solidify an aqueous solution after bridging, achieving a high ice packing factor (IPF). First, experiments using small-scale test cells have been conducted to confirm the present idea, and then we have performed experiments using a large vessel with an early practical size. It was seen that a large pressure increase is prevented for the initial concentration of the solution C0 of 1.0%, and IPF obtained using the solution is much greater than 0.65 using pure water for which the solidification must be stopped before the bridging.

What is cerebral small vessel disease?

International Nuclear Information System (INIS)

Onodera, Osamu

2011-01-01

An accumulating amount of evidence suggests that the white matter hyperintensities on T 2 weighted brain magnetic resonance imaging predict an increased risk of dementia and gait disturbance. This state has been proposed as cerebral small vessel disease, including leukoaraiosis, Binswanger's disease, lacunar stroke and cerebral microbleeds. However, the concept of cerebral small vessel disease is still obscure. To understand the cerebral small vessel disease, the precise structure and function of cerebral small vessels must be clarified. Cerebral small vessels include several different arteries which have different anatomical structures and functions. Important functions of the cerebral small vessels are blood-brain barrier and perivasucular drainage of interstitial fluid from the brain parenchyma. Cerebral capillaries and glial endfeet, take an important role for these functions. However, the previous pathological investigations on cerebral small vessels have focused on larger arteries than capillaries. Therefore little is known about the pathology of capillaries in small vessel disease. The recent discoveries of genes which cause the cerebral small vessel disease indicate that the cerebral small vessel diseases are caused by a distinct molecular mechanism. One of the pathological findings in hereditary cerebral small vessel disease is the loss of smooth muscle cells, which is an also well-recognized finding in sporadic cerebral small vessel disease. Since pericytes have similar character with the smooth muscle cells, the pericytes should be investigated in these disorders. In addition, the loss of smooth muscle cells may result in dysfunction of drainage of interstitial fluid from capillaries. The precise correlation between the loss of smooth muscle cells and white matter disease is still unknown. However, the function that is specific to cerebral small vessel may be associated with the pathogenesis of cerebral small vessel disease. (author)

Attitudes and Experiences of Close Interethnic Friendships Among Native Emerging Adults: A Mixed-Methods Investigation

OpenAIRE

Jones, Merrill L.

2017-01-01

This study included 114 Native adults and 6 Native/non-Native pairs of friends (age 18-25). Experiences and attitudes for close interethnic friendships were investigated. Friendship patterns and predictors were quantitatively assessed for the 114 Natives, with qualitative examination of the development and qualities of the six friend pairs. Results of quantitative analysis revealed that 80% of this sample reported friendship investment with Whites, and 55% reported friendship investment wi...

UO2 corrosion in high surface-area-to-volume batch experiments

International Nuclear Information System (INIS)

Bates, J. K.; Finch, R. J.; Hanchar, J. M.; Wolf, S. F.

1997-01-01

Unsaturated drip tests have been used to investigate the alteration of unirradiated UO 2 and spent UO 2 fuel in an unsaturated environment such as may be expected in the proposed repository at Yucca Mountain. In these tests, simulated groundwater is periodically injected onto a sample at 90 C in a steel vessel. The solids react with the dripping groundwater and water condensed on surfaces to form a suite of U(VI) alteration phases. Solution chemistry is determined from leachate at the bottom of each vessel after the leachate stops interacting with the solids. A more detailed knowledge of the compositional evolution of the leachate is desirable. By providing just enough water to maintain a thin film of water on a small quantity of fuel in batch experiments, we can more closely monitor the compositional changes to the water as it reacts to form alteration phases

In-vessel coolability and steam explosion in Nordic BWRs

Energy Technology Data Exchange (ETDEWEB)

Ma, W.; Li, L.; Hansson, R.; Villanueva, W.; Kudinov, P.; Manickam, L.; Tran, C.-T. (Royal Institute of Technology (KTH) (Sweden))

2011-05-15

The objective of this research is to reduce the uncertainty in quantification of steam explosion risk and in-vessel coolability in the Nordic BWR plants which employ cavity flooding as severe accident management (SAM) strategy. To quantify the coolability of debris bed packed with irregular particles, the friction laws of fluid flow in particulate beds packed with non-spherical particles were investigated on the POMECO-FL test facility, and the experimental data suggest that the Ergun equation is applicable if the effective particle diameter of the particles is represented by the equivalent diameter of the particles, which is the product of Sauter mean diameter and shape factor of the particles. One-way coupling analysis between PECM model for melt pool heat transfer and ANSYS thermo-structural mechanics was performed to analyze the vessel creep, and the results revealed two different modes of vessel failure: a 'ballooning' of the vessel bottom and a 'localized creep' concentrated within the vicinity of the top surface of the melt pool. Single-droplet steam explosion experiments were carried out by using oxidic mixture of WO{sub 3}-CaO, and the results show an apparent difference in steam explosion energetics between the eutectic and non-eutectic melts at low melt superheat (100 deg. C). (Author)

Vacuum vessel for the tandem Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Gerich, J.W.

1986-01-01

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

In-vessel coolability and steam explosion in Nordic BWRs

International Nuclear Information System (INIS)

Ma, W.; Li, L.; Hansson, R.; Villanueva, W.; Kudinov, P.; Manickam, L.; Tran, C.-T.

2011-05-01

The objective of this research is to reduce the uncertainty in quantification of steam explosion risk and in-vessel coolability in the Nordic BWR plants which employ cavity flooding as severe accident management (SAM) strategy. To quantify the coolability of debris bed packed with irregular particles, the friction laws of fluid flow in particulate beds packed with non-spherical particles were investigated on the POMECO-FL test facility, and the experimental data suggest that the Ergun equation is applicable if the effective particle diameter of the particles is represented by the equivalent diameter of the particles, which is the product of Sauter mean diameter and shape factor of the particles. One-way coupling analysis between PECM model for melt pool heat transfer and ANSYS thermo-structural mechanics was performed to analyze the vessel creep, and the results revealed two different modes of vessel failure: a 'ballooning' of the vessel bottom and a 'localized creep' concentrated within the vicinity of the top surface of the melt pool. Single-droplet steam explosion experiments were carried out by using oxidic mixture of WO 3 -CaO, and the results show an apparent difference in steam explosion energetics between the eutectic and non-eutectic melts at low melt superheat (100 deg. C). (Author)

Stress analysis of pressure vessels in nuclear power plants: pt.2

International Nuclear Information System (INIS)

Kim, C.W.; Chu, Y.W.

1976-01-01

Stress analysis of tapered cylinder of reactor vessels in investigated by means of the intersection method. The tapered cylinder is approximated into three models average cylinder, conical frustum, and ring. The results are compared with those of the finite element method program and an experiment. In this paper, the following results are obtained: (1) the best approximation has been obtained by the ring model analysis: (2) the intersection analysis of the tapered cylinder by the ring model shows a sufficient accuracy for the stress analysis of reactor vessels. (author)

Recent evaluation of 'wet' thermal annealing to resolve reactor pressure vessel embrittlement

International Nuclear Information System (INIS)

Server, W.L.; Biemiller, E.C.

1993-01-01

Prior to the decision to close the Yankee Rowe plant in 1992, a great deal of effort was expended in trying to resolve the degree of neutron embrittlement that the reactor pressure vessel had experienced after 30 years of operation. One mitigative measure that was examined in detail was the possibility of performing a relatively low temperature thermal anneal (at approximately 650 deg. F) to partially restore the original design level of mechanical properties of the reactor pressure vessel beltline region which were lost due to the neutron radiation exposure. This low temperature anneal was to involve heating of the primary coolant water using pump heat in a similar manner as that used to anneal the Belgian BR-3 reactor pressure vessel in the early 1980s. This 'wet' anneal was successful in recovering mechanical properties for the BR-3 vessel, but the extent of the recovery, as well as the rate of re-embrittlement after the anneal, were issues that were difficult to quantify since the exact reactor pressure vessel steels were not available for experimental verification. For the case of Yankee Rowe, material was available from past surveillance programs for at least one of the materials in the vessel, as well as materials obtained from various sources which could act as bounding surrogates. An irradiation /annealing/reirradiation program was developed to better quantify the degree of recovery and re-embrittlement for these materials, but this program was halted before significant test results were obtained. Prior to the initiation of the testing program, a review of past annealing data was performed and the data were scrutinized for direct relevance to the annealing response of the Yankee Rowe vessel. This paper discusses the results derived from this review. The results from the critical review of the past annealing data indicated that a 'wet' anneal of the Yankee Rowe vessel may have been successful in reducing the degree of embrittlement to the point that the

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

Radiology trainer. Torso, internal organs and vessels. 2. ed.

International Nuclear Information System (INIS)

Staebler, Axel; Erlt-Wagner, Birgit

2013-01-01

The radiology training textbook is based on case studies of the clinical experience, including radiological imaging and differential diagnostic discussion. The scope of this volume covers the torso, internal organs and vessels. The following issues are discussed: lungs, pleura, mediastinum; heart and vascular system; upper abdomen organs; gastrointestinal tract; urogenital system.

Heat dissipation research on the water-cooling channel of HL-2M in-vessel coils

Energy Technology Data Exchange (ETDEWEB)

Jiang, J., E-mail: jiangjiaming@swip.ac.cn; Liu, Y.; Chen, Q.; Ji, X.Q.

2017-04-15

Highlights: • The joule heat of in-vessel coils is very difficult to dissipate inside HL-2M vacuum vessel. • Heat dissipation model of the coil includes the joule heat model, the heat conduction model and the heat transfer model. • The CFD analysis has been done for the coil-water cooling, with comparison with the date of theoretical analysis and experiment. • The result shows water-cooling channel is good for the joule heat transfer and taken away. - Abstract: HL-2M in-vessel coils are positioned in high vacuum circumstance, and they will generate joule heat when they carry 15 kA electrical current, but joule heat is very difficult to dissipate in vacuum, so a hollow cable with 8 mm inner diameter is design as water-cooling channel for heat convection. By using the methods of the theoretical derivation, together with CFD numeric simulation method and the experiment of the heat transfer, the water channel of HL-2M in-vessel coils has been studied, and the temperature of HL-2M in-vessel coils under different cooling water flow rates is obtained and acceptable. Simultaneously, the external cooling water supply system parameters for the water-cooling channel of the coils are estimated. Three methods’ results are in good agreement; the theoretical model is verified and could be popularized for predicting the temperature rise of HL-2M in-vessel coils.

Measurement of flow velocity fields in small vessel-mimic phantoms and vessels of small animals using micro ultrasonic particle image velocimetry (micro-EPIV).

Science.gov (United States)

Qian, Ming; Niu, Lili; Wang, Yanping; Jiang, Bo; Jin, Qiaofeng; Jiang, Chunxiang; Zheng, Hairong

2010-10-21

Determining a multidimensional velocity field within microscale opaque fluid flows is needed in areas such as microfluidic devices, biofluid mechanics and hemodynamics research in animal studies. The ultrasonic particle image velocimetry (EchoPIV) technique is appropriate for measuring opaque flows by taking advantage of PIV and B-mode ultrasound contrast imaging. However, the use of clinical ultrasound systems for imaging flows in small structures or animals has limitations associated with spatial resolution. This paper reports on the development of a high-resolution EchoPIV technique (termed as micro-EPIV) and its application in measuring flows in small vessel-mimic phantoms and vessels of small animals. Phantom experiments demonstrate the validity of the technique, providing velocity estimates within 4.1% of the analytically derived values with regard to the flows in a small straight vessel-mimic phantom, and velocity estimates within 5.9% of the computationally simulated values with regard to the flows in a small stenotic vessel-mimic phantom. Animal studies concerning arterial and venous flows of living rats and rabbits show that the micro-EPIV-measured peak velocities within several cardiac cycles are about 25% below the values measured by the ultrasonic spectral Doppler technique. The micro-EPIV technique is able to effectively measure the flow fields within microscale opaque fluid flows.

Measurement of flow velocity fields in small vessel-mimic phantoms and vessels of small animals using micro ultrasonic particle image velocimetry (micro-EPIV)

International Nuclear Information System (INIS)

Qian Ming; Niu Lili; Jiang Bo; Jin Qiaofeng; Jiang Chunxiang; Zheng Hairong; Wang Yanping

2010-01-01

Determining a multidimensional velocity field within microscale opaque fluid flows is needed in areas such as microfluidic devices, biofluid mechanics and hemodynamics research in animal studies. The ultrasonic particle image velocimetry (EchoPIV) technique is appropriate for measuring opaque flows by taking advantage of PIV and B-mode ultrasound contrast imaging. However, the use of clinical ultrasound systems for imaging flows in small structures or animals has limitations associated with spatial resolution. This paper reports on the development of a high-resolution EchoPIV technique (termed as micro-EPIV) and its application in measuring flows in small vessel-mimic phantoms and vessels of small animals. Phantom experiments demonstrate the validity of the technique, providing velocity estimates within 4.1% of the analytically derived values with regard to the flows in a small straight vessel-mimic phantom, and velocity estimates within 5.9% of the computationally simulated values with regard to the flows in a small stenotic vessel-mimic phantom. Animal studies concerning arterial and venous flows of living rats and rabbits show that the micro-EPIV-measured peak velocities within several cardiac cycles are about 25% below the values measured by the ultrasonic spectral Doppler technique. The micro-EPIV technique is able to effectively measure the flow fields within microscale opaque fluid flows.

Structural features and in-service inspection of the LTHR-200 pressure vessel

International Nuclear Information System (INIS)

Xiong Dunshi; He Shuyan; Liu Junjie; Yu Suyuan

1993-01-01

LTHR-200 is a low temperature district-heating reactor. It adopts double-shell design pressure vessel and metal containment. Because of the safety and structural features of the reactor, the in-service inspection of the pressure vessel can be simplified greatly. LTHR-200 is an integrated arrangement. Both its core components and the main heat exchangers are contained in the reactor pressure vessel. The coolant of the main loop is run by a full-power natural circulation and there need no main pumps and pipes. Thus, the reactor pressure vessel constitutes the pressure boundary of the reactor's main loop coolant. In regard to these features, a small-sized containment is designed for the reactor. The metal safety container with a small volume is placed closely around the reactor pressure vessel. Outside the metal containment, there is a large reinforced concrete construction for the reactor. Their main operation and design parameters are as follows: The pressure vessel: operation pressure = 2.4 MPa; design pressure = 3.0 MPa; design temperature = 250 deg C; 40 year fast neutron (E>1MeV) fluence in the belt-line region = < 10E16n/cm; internal diameter = 5000 mm; material SA516-70; shell thickness 65 mm; The metal containment: maximum operation pressure = 1.8 MPa; design pressure = 1.8 MPa; design temperature = 250 deg. C; upper internal diameter 7000 mm; lower internal diameter = 5600 mm; material = SA516-70; shell thickness, upper part = 80 mm; lower part = 50 mm. All penetrating pipes through the pressure vessel are located at the top penetration section of the shell. All the internal diameters of penetrating pipes are less than 50 mm. Inside and outside the metal containment wall respectively, isolating valves are connected to the reactor coolant pipe which passes through the containment. These two isolating valves use different driving methods. Every penetrating part of the reactor construction uses a proper form of structure according to safety requirements

Flexible Composite-Material Pressure Vessel

Science.gov (United States)

Brown, Glen; Haggard, Roy; Harris, Paul A.

2003-01-01

A proposed lightweight pressure vessel would be made of a composite of high-tenacity continuous fibers and a flexible matrix material. The flexibility of this pressure vessel would render it (1) compactly stowable for transport and (2) more able to withstand impacts, relative to lightweight pressure vessels made of rigid composite materials. The vessel would be designed as a structural shell wherein the fibers would be predominantly bias-oriented, the orientations being optimized to make the fibers bear the tensile loads in the structure. Such efficient use of tension-bearing fibers would minimize or eliminate the need for stitching and fill (weft) fibers for strength. The vessel could be fabricated by techniques adapted from filament winding of prior composite-material vessels, perhaps in conjunction with the use of dry film adhesives. In addition to the high-bias main-body substructure described above, the vessel would include a low-bias end substructure to complete coverage and react peak loads. Axial elements would be overlaid to contain damage and to control fiber orientation around side openings. Fiber ring structures would be used as interfaces for connection to ancillary hardware.

Development of design Criteria for ITER In-vessel Components

International Nuclear Information System (INIS)

Sannazzaro, G.; Barabash, V.; Kang, S.C.; Fernandez, E.; Kalinin, G.; Obushev, A.; Martínez, V.J.; Vázquez, I.; Fernández, F.; Guirao, J.

2013-01-01

Absrtract: The components located inside the ITER vacuum chamber (in-vessel components – IC), due to their specific nature and the environments they are exposed to (neutron radiation, high heat fluxes, electromagnetic forces, etc.), have specific design criteria which are, in this paper, referred as Structural Design Criteria for In-vessel Components (SDC-IC). The development of these criteria started in the very early phase of the ITER design and followed closely the criteria of the RCC-MR code. Specific rules to include the effect of neutron irradiation were implemented. In 2008 the need of an update of the SDC-IC was identified to add missing specifications, to implement improvements, to modernise rules including recent evolutions in international codes and regulations (i.e. PED). Collaboration was set up between ITER Organization (IO), European (EUDA) and Russian Federation (RFDA) Domestic Agencies to generate a new version of SDC-IC. A Peer Review Group (PRG) composed by members of the ITER Organization and all ITER Domestic Agencies and code experts was set-up to review the proposed modifications, to provide comments, contributions and recommendations

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

Percutaneous Retrieval of Foreign Bodies Around Vital Vessels Aided with Vascular Intervention: A Technical Note

Energy Technology Data Exchange (ETDEWEB)

Yang, Xiu-Jun, E-mail: woothingyang2008@126.com [Shanghai Eighth People’s Hospital, Department of Radiology (China); Xing, Guang-Fu, E-mail: xgf8848@126.com [Shanghai Eighth People’s Hospital, Department of General Surgery (China)

2015-10-15

ObjectiveTo describe a new interventional technique to remove foreign bodies (FBs) embedded in soft tissues around vital vessels.MethodsUnder fluoroscopic guidance and using local anesthesia, percutaneous removal of FBs was performed using forceps in nine patients. All patients suffered from a metallic soft tissue FB located in close proximity to important vessels and one also had a small traumatic pseudoaneurysm adjacent to the FB. Prior to removal of the FB, the position of the nearest vessel was identified using a guide wire or catheter placed into the vessel. Balloon catheter was also simultaneously used to temporarily stop the blood flow of the nearest artery during the FB removal in three of the nine patients.ResultsAll of the nine FBs with 0–2 mm interval to the nearest vessel were successfully removed in the nine patients without any serious complications. The removed FBs measured 3–12 mm in length and 1–3 mm in width. The total fluoroscopic time of retrieval of each FB was 5–9 min (mean, 6.4 min). The volume of intraoperative bleeding ranged from 5 to 12 ml (mean, 7.5 ml). The length of hospital stay for each patient ranged from 4 to 8 days (mean, 5.5 days).ConclusionVascular intervention-aided percutaneous FB removal is minimally invasive and an effective method for removal of FBs around vital vessels.

Large inelastic deformation analysis of steel pressure vessels at high temperature

International Nuclear Information System (INIS)

Ikonen, K.

2001-01-01

This publication describes the calculation methodology developed for a large inelastic deformation analysis of pressure vessels at high temperature. Continuum mechanical formulation related to a large deformation analysis is presented. Application of the constitutive equations is simplified when the evolution of stress and deformation state of an infinitesimal material element is considered in the directions of principal strains determined by the deformation during a finite time increment. A quantitative modelling of time dependent inelastic deformation is applied for reactor pressure vessel steels. Experimental data of uniaxial tensile, relaxation and creep tests performed at different laboratories for reactor pressure vessel steels are investigated and processed. An inelastic deformation rate model of strain hardening type is adopted. The model simulates well the axial tensile, relaxation and creep tests from room temperature to high temperature with only a few fitting parameters. The measurement data refined for the inelastic deformation rate model show useful information about inelastic deformation phenomena of reactor pressure vessel steels over a wide temperature range. The methodology and calculation process are validated by comparing the calculated results with measurements from experiments on small scale pressure vessels. A reasonably good agreement, when taking several uncertainties into account, is obtained between the measured and calculated results concerning deformation rate and failure location. (orig.)

Reliability analysis of pipelines and pressure vessels at nuclear power plants

International Nuclear Information System (INIS)

Klemin, A.I.; Shiverskij, E.A.

1979-01-01

Reliability analysis of pipelines and pressure vessels at NPP is given. The main causes and failure mechanisms of these elements, the ways of reliability improvement and preventing of great damages are considered. The reliability estimation methods both according to the statistical operation data and under the conditions of absence of failure statistics are given. The main characteristics and actual reliability factors of pipelines and pressure vessels of three home NPP: the first in the world NPP, VK-50 and Beloyarsk NPP, are presented. From the start-up there were practically no failures of the pipelines and pressure vessels at the VK-50 pilot installation. The analysis of the operation experience of the first and second blocks of the Beloyarsk NPP, as well as the first in the world NPP, shows that the most part of failures of the pipelines and pressure vessels of these energy blocks with the channel reactors is connected with the coolant leakage at minority pipelines of a small diameter. The most part of failures at individual pipelines of the first and second blocks of the Beloyarsk NPP are connected with the leakages of stuffing boxes of switching off devices. It is noted that serious failures of large pipelines and pressure vessels at all home NPP under operation have not been observed

Oxidation effects during corium melt in-vessel retention

Energy Technology Data Exchange (ETDEWEB)

Almyashev, V.I.; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Sulatsky, A.A.; Vitol, S.A. [Alexandrov Scientific-Research Institute of Technology (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V. [Ioffe Institute, St. Petersburg (Russian Federation); Bechta, S. [Royal Institute of Technology (KHT), Stockholm (Sweden); Barrachin, M.; Fichot, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), St Paul lez Durance (France); Bottomley, P.D., E-mail: paul.bottomley@ec.europa.eu [Joint Research Centre, Institut für Transurane (ITU), Karlsruhe (Germany); Fischer, M. [AREVA GmbH, Erlangen (Germany); Piluso, P. [CEA Cadarache-DEN/DTN/STRI (France)

2016-08-15

Highlights: • Corium–steel interaction tests were re-examined particularly for transient processes. • Oxidation of corium melt was sensitive to oxidant supply and surface characteristics. • Consequences for vessel steel corrosion rates in severe accidents were discussed. - Abstract: In the in-vessel corium retention studies conducted on the Rasplav-3 test facility within the ISTC METCOR-P project and OECD MASCA program, experiments were made to investigate transient processes taking place during the oxidation of prototypic molten corium. Qualitative and quantitative data have been produced on the sensitivity of melt oxidation rate to the type of oxidant, melt composition, molten pool surface characteristics. The oxidation rate is a governing factor for additional heat generation and hydrogen release; also for the time of secondary inversion of oxidic and metallic layers of corium molten pool.

Acrylic vessel cleaning tests

International Nuclear Information System (INIS)

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

1997-01-01

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

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

Development of PWR pressure vessel steels

International Nuclear Information System (INIS)

Druce, S.; Edwards, B.

1982-01-01

Requirements to be met by vessel steels for pressurized water reactors are analyzed. Chemicat composition of low-alloyed steels, mechanical properties of sheets and forgings made of these steels and changes in the composition and properties over the wall thickness of the reactor vessel are presented. Problems of the vessel manufacturing including welding and heat treatment processes of sheets and forgings are considered. Special attention is paid to steel embrittlement during vessel fabrication and operation (radiation embrittlement, thermal embrittlement). The role of non-metal inclusions and their effect on anisotropy of fracture toughness is discussed. Possible developments of vessel steels and procedures for producing reactor vessels are reviewed

Development of PWR pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

Druce, S.; Edwards, B.

1982-01-01

Requirements to be met by vessel steels for pressurized water reactors are analyzed. Chemicat composition of low-alloyed steels, mechanical properties of sheets and forgings made of these steels and changes in the composition and properties over the wall thickness of the reactor vessel are presented. Problems of the vessel manufacturing including welding and heat treatment processes of sheets and forgings are considered. Special attention is paid to steel embrittlement during vessel fabrication and operation (radiation embrittlement, thermal embrittlement). The role of non-metal inclusions and their effect on anisotropy of fracture toughness is discussed. Possible developments of vessel steels and procedures for producing reactor vessels are reviewed.

Development of glia and blood vessels in the internal capsule of rats.

Science.gov (United States)

Earle, K L; Mitrofanis, J

1998-02-01

microglia are seen in the thalamus. In summary, our results indicate that all three types of glia in the internal capsule are associated closely with the vasculature, suggesting they may play a role in the development of the blood-brain barrier among the vessels in this white matter region of the forebrain.

Clay Corner: Recreating Chinese Bronze Vessels.

Science.gov (United States)

Gamble, Harriet

1998-01-01

Presents a lesson where students make faux Chinese bronze vessels through slab or coil clay construction after they learn about the history, function, and design of these vessels. Utilizes a variety of glaze finishes in order to give the vessels an aged look. Gives detailed guidelines for creating the vessels. (CMK)

Gammatography of thick lead vessels

International Nuclear Information System (INIS)

Raghunath, V.M.; Bhatnagar, P.K.; Sundaram, V.M.

1979-01-01

Radiography, scintillation and GM counting and dose measurements using ionisation chamber equipment are commonly used for detecting flaws/voids in materials. The first method is mostly used for steel vessels and to a lesser extent thin lead vessels also and is essentially qualitative. Dose measuring techniques are used for very thick and large lead vessels for which high strength radioactive sources are required, with its inherent handling problems. For vessels of intermediate thicknesses, it is ideal to use a small strength source and a GM or scintillation counter assembly. At the Reactor Research Centre, Kalpakkam, such a system was used for checking three lead vessels of thicknesses varying from 38mm to 65mm. The tolerances specified were +- 4% variation in lead thickness. The measurements also revealed the non concentricity of one vessel which had a thickness varying from 38mm to 44mm. The second vessel was patently non-concentric and the dimensional variation was truly reproduced in the measurements. A third vessel was fabricated with careful control of dimensions and the measurements exhibited good concentricity. Small deviations were observed, attributable to imperfect bondings between steel and lead. This technique has the following advantages: (a) weaker sources used result in less handling problems reducing the personnel exposures considerably; (b) the sensitivity of the instrument is quite good because of better statistics; (c) the time required for scanning a small vessel is more, but a judicious use of a scintillometer for initial fast scan will help in reducing the total scanning time; (d) this method can take advantage of the dimensional variations themselves to get the calibration and to estimate the deviations from specified tolerances. (auth.)

Smooth muscle cell recruitment to lymphatic vessels requires PDGFB and impacts vessel size but not identity.

Science.gov (United States)

Wang, Yixin; Jin, Yi; Mäe, Maarja Andaloussi; Zhang, Yang; Ortsäter, Henrik; Betsholtz, Christer; Mäkinen, Taija; Jakobsson, Lars

2017-10-01

Tissue fluid drains through blind-ended lymphatic capillaries, via smooth muscle cell (SMC)-covered collecting vessels into venous circulation. Both defective SMC recruitment to collecting vessels and ectopic recruitment to lymphatic capillaries are thought to contribute to vessel failure, leading to lymphedema. However, mechanisms controlling lymphatic SMC recruitment and its role in vessel maturation are unknown. Here, we demonstrate that platelet-derived growth factor B (PDGFB) regulates lymphatic SMC recruitment in multiple vascular beds. PDGFB is selectively expressed by lymphatic endothelial cells (LECs) of collecting vessels. LEC-specific deletion of Pdgfb prevented SMC recruitment causing dilation and failure of pulsatile contraction of collecting vessels. However, vessel remodelling and identity were unaffected. Unexpectedly, Pdgfb overexpression in LECs did not induce SMC recruitment to capillaries. This was explained by the demonstrated requirement of PDGFB extracellular matrix (ECM) retention for lymphatic SMC recruitment, and the low presence of PDGFB-binding ECM components around lymphatic capillaries. These results demonstrate the requirement of LEC-autonomous PDGFB expression and retention for SMC recruitment to lymphatic vessels, and suggest an ECM-controlled checkpoint that prevents SMC investment of capillaries, which is a common feature in lymphedematous skin. © 2017. Published by The Company of Biologists Ltd.

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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)

Numerical investigations on pressurized AL-composite vessel response to hypervelocity impacts: Comparison between experimental works and a numerical code

Directory of Open Access Journals (Sweden)

Mespoulet Jérôme

2015-01-01

Full Text Available Response of pressurized composite-Al vessels to hypervelocity impact of aluminum spheres have been numerically investigated to evaluate the influence of initial pressure on the vulnerability of these vessels. Investigated tanks are carbon-fiber overwrapped prestressed Al vessels. Explored internal air pressure ranges from 1 bar to 300 bar and impact velocity are around 4400 m/s. Data obtained from experiments (Xray radiographies, particle velocity measurement and post-mortem vessels have been compared to numerical results given from LS-DYNA ALE-Lagrange-SPH full coupling models. Simulations exhibit an under estimation in term of debris cloud evolution and shock wave propagation in pressurized air but main modes of damage/rupture on the vessels given by simulations are coherent with post-mortem recovered vessels from experiments. First results of this numerical work are promising and further simulation investigations with additional experimental data will be done to increase the reliability of the simulation model. The final aim of this crossed work is to numerically explore a wide range of impact conditions (impact angle, projectile weight, impact velocity, initial pressure that cannot be explore experimentally. Those whole results will define a rule of thumbs for the definition of a vulnerability analytical model for a given pressurized vessel.

Streamlined vessels for speedboats: Macro modifications of shark skin design applications

Science.gov (United States)

Ibrahim, M. D.; Amran, S. N. A.; Zulkharnain, A.; Sunami, Y.

2018-01-01

Functional properties of shark denticles have caught the attention of engineers and scientist today due to the hydrodynamic effects of its skin surface roughness. The skin of a fast swimming shark reveals riblet structures that help to reduce skin friction drag, shear stresses, making its movement to be more efficient and faster. Inspired by the structure of the shark skin denticles, our team has conducted a study on alternative on improving the hydrodynamic design of marine vessels by applying the simplified version of shark skin skin denticles on the surface hull of the vessels. Models used for this study are constructed and computational fluid dynamic (CFD) simulations are then carried out to predict the effectiveness of the hydrodynamic effects of the biomimetic shark skins on those models. Interestingly, the numerical calculated results obtained shows that the presence of biomimetic shark skin implemented on the vessels give improvements in the maximum speed as well as reducing the drag force experience by the vessels. The pattern of the wave generated post cruising area behind the vessels can also be observed to reduce the wakes and eddies. Theoretically, reduction of drag force provides a more efficient vessel with a better cruising speed. To further improve on this study, the authors are now actively arranging an experimental procedure in order to verify the numerical results obtained by CFD. The experimental test will be carried out using an 8 metre flow channel provided by University Malaysia Sarawak, Malaysia.

The experience of being a middle-aged close relative of a person who has suffered a stroke--six months after discharge from a rehabilitation clinic.

Science.gov (United States)

Bäckström, Britt; Sundin, Karin

2010-03-01

Being a close relative brings with it a large number of consequences, with the life situation changing over time. The aim of this study was to illuminate the experiences of being a middle-aged close relative of a person who has suffered a stroke 6 months after being discharged from a medical rehabilitation clinic. Narrative interviews were conducted with nine middle-aged close relatives and analysed using a content analysis with a latent approach. The analysis revealed that being close to someone who had suffered a stroke 6 months after discharge meant; a struggling for control and a renewal of family life in the shadow of suffering and hope. The middle-aged close relatives began to perceive the changed reality. They were struggling to take on something new, become reconciled and find a balance in their family life. Their ability to work, relief from caring concerns and having support and togetherness with others seemed to be essential for the close relatives in their efforts to manage their life situation and maintain their well-being. Having reached the 'halfway point' in their lives and still with half of their life in front of them created worries. They felt dejected about their changed relationships and roles, experience a sense of loss of shared child responsibilities, a negative impact on their marital relationships and sexual satisfaction. They felt trapped in a caring role and they worried about how to endure in the future. The middle-aged close relatives' experiences were of being alone and neglected, in an arduous and complex life situation filled with loss and grief. The findings highlights that health professionals need to see and listen to the close relatives' experiences of transition in order to provide appropriate support adjusted to their varying needs during a time of renewal.

In-Vessel Coil Material Failure Rate Estimates for ITER Design Use

Energy Technology Data Exchange (ETDEWEB)

L. C. Cadwallader

2013-01-01

The ITER international project design teams are working to produce an engineering design for construction of this large tokamak fusion experiment. One of the design issues is ensuring proper control of the fusion plasma. In-vessel magnet coils may be needed for plasma control, especially the control of edge localized modes (ELMs) and plasma vertical stabilization (VS). These coils will be lifetime components that reside inside the ITER vacuum vessel behind the blanket modules. As such, their reliability is an important design issue since access will be time consuming if any type of repair were necessary. The following chapters give the research results and estimates of failure rates for the coil conductor and jacket materials to be used for the in-vessel coils. Copper and CuCrZr conductors, and stainless steel and Inconel jackets are examined.

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

International Nuclear Information System (INIS)

Domian, H.A.

1987-02-01

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

Close encounters: an examination of UFO experiences.

Science.gov (United States)

Spanos, N P; Cross, P A; Dickson, K; DuBreuil, S C

1993-11-01

Ss who reported UFO experiences were divided into those whose experiences were nonintense (e.g., seeing lights and shapes in the sky) and those whose experiences were intense (e.g., seeing and communicating with aliens or missing time). On a battery of objective tests Ss in these 2 groups did not score as more psychopathological, less intelligent, or more fantasy prone and hypnotizable than a community comparison group or a student comparison group. However, Ss in the UFO groups believed more strongly in space alien visitation than did comparison Ss. The UFO experiences of Ss in the intense group were more frequently sleep-related than the experiences of Ss in the nonintense group. Among the combined UFO Ss, intensity of UFO experiences correlated significantly with inventories that assessed proneness toward fantasy and unusual sensory experiences. Implications are discussed.

Segmentation and packaging reactor vessels internals

International Nuclear Information System (INIS)

Boucau, Joseph

2014-01-01

Document available in abstract form only, full text follows: With more than 25 years of experience in the development of reactor vessel internals and reactor vessel segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since disposal cost is a key factor, it is important to plan and optimize waste segmentation and packaging. The choice of the optimum cutting technology is also important for a successful project implementation and depends on some specific constraints. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. The usual method is to start at the end of the process, by evaluating handling of the containers, the waste disposal requirements, what type and size of containers are available for the different disposal options, and working backwards to select a cutting method and finally the cut geometry required. The 3-D models can include intelligent data such as weight, center of gravity, curie content, etc, for each segmented piece, which is very useful when comparing various cutting, handling and packaging options. The detailed 3-D analyses and thorough characterization assessment can draw the attention to material potentially subject to clearance, either directly or after certain period of decay, to allow recycling and further disposal cost reduction. Westinghouse has developed a variety of special cutting and handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a successful reactor vessel internals

Americium behaviour in plastic vessels

Energy Technology Data Exchange (ETDEWEB)

Legarda, F.; Herranz, M. [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain); Idoeta, R., E-mail: raquel.idoeta@ehu.e [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain); Abelairas, A. [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain)

2010-07-15

The adsorption of {sup 241}Am dissolved in water in different plastic storage vessels was determined. Three different plastics were investigated with natural and distilled waters and the retention of {sup 241}Am by these plastics was studied. The same was done by varying vessel agitation time, vessel agitation speed, surface/volume ratio of water in the vessels and water pH. Adsorptions were measured to be between 0% and 70%. The adsorption of {sup 241}Am is minimized with no water agitation, with PET or PVC plastics, and by water acidification.

Americium behaviour in plastic vessels

International Nuclear Information System (INIS)

Legarda, F.; Herranz, M.; Idoeta, R.; Abelairas, A.

2010-01-01

The adsorption of 241 Am dissolved in water in different plastic storage vessels was determined. Three different plastics were investigated with natural and distilled waters and the retention of 241 Am by these plastics was studied. The same was done by varying vessel agitation time, vessel agitation speed, surface/volume ratio of water in the vessels and water pH. Adsorptions were measured to be between 0% and 70%. The adsorption of 241 Am is minimized with no water agitation, with PET or PVC plastics, and by water acidification.

PWR vessel flaw distribution development

International Nuclear Information System (INIS)

Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

1990-01-01

This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

Baking of SST-1 vacuum vessel modules and sectors

International Nuclear Information System (INIS)

Pathan, Firozkhan S; Khan, Ziauddin; Yuvakiran, Paravastu; George, Siju; Ramesh, Gattu; Manthena, Himabindu; Shah, Virendrakumar; Raval, Dilip C; Thankey, Prashant L; Dhanani, Kalpesh R; Pradhan, Subrata

2012-01-01

SST-1 Tokamak is a steady state super-conducting tokamak for plasma discharge of 1000 sec duration. The plasma discharge of such long time duration can be obtained by reducing the impurities level, which will be possible only when SST-1 vacuum chamber is pumped to ultra high vacuum. In order to achieve UHV inside the chamber, the baking of complete vacuum chamber has to be carried out during pumping. For this purpose the C-channels are welded inside the vacuum vessel. During baking of vacuum vessel, these welded channels should be helium leak tight. Further, these U-channels will be in accessible under operational condition of SST-1. So, it will not possible to repair if any leak is developed during experiment. To avoid such circumstances, a dedicated high vacuum chamber is used for baking of the individual vacuum modules and sectors before assembly so that any fault during welding of the channels will be obtained and repaired. This paper represents the baking of vacuum vessel modules and sectors and their temperature distribution along the entire surface before assembly.

Heat and mass transfer in a concrete pressure vessel

International Nuclear Information System (INIS)

Zangle, K.; Sadouki, H.; Wittmann, F.H.

1989-01-01

Pressure vessels of prestressed concrete for high temperature reactors are subjected to high mechanical and thermal stresses during the reactors normal working conditions and in particular accidental conditions. According to a large temperature gradient between the inner liner and the outer side of the thickwalled vessel, physical as well as chemical processes take place in concrete. Temperature and moisture content of concrete have a big influence on these processes. During the last years different investigations have been conducted in order to determine characteristic values of concrete under these conditions. At present the authors conduct a series of experiments on model vessels of prestressed concrete and a large number of small specimens. The aims of these tests can be briefly summarized as follows: experimental determination of transport coefficients for a numerical analysis; determination of chemical reactions under hydrothermal conditions and their significance for the risk of corrosion; determination of temperature and moisture distribution as a function of time; and determination of the strength development in the zones subjected to elevated temperatures

Pressurized-thermal-shock experiments

International Nuclear Information System (INIS)

Whitman, G.D.; McCulloch, R.W.

1982-01-01

The primary objective of the ORNL pressurized-thermal-shock (PTS) experiments is to verify analytical methods that are used to predict the behavior of pressurized-water-reactor vessels under these accident conditions involving combined pressure and thermal loading. The criteria on which the experiments are based are: scale large enough to attain effective flaw border triaxial restraint and a temperature range sufficiently broad to produce a progression from frangible to ductile behavior through the wall at a given time; use of materials that can be completely characterized for analysis; stress states comparable to the actual vessel in zones of potential flaw extension; range of behavior to include cleavage initiation and arrest, cleavage initiation and arrest on the upper shelf, arrest in a high K/sub I/ gradient, warm prestressing, and entirely ductile behavior; long and short flaws with and without stainless steel cladding; and control of loads to prevent vessel burst, except as desired. A PTS test facility is under construction which will enable the establishment and control of wall temperature, cooling rate, and pressure on an intermediate test vessel (ITV) in order to simulate stress states representative of an actual reactor pressure vessel

Hydrogen storage in insulated pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Aceves, S.M.; Garcia-Villazana, O. [Lawrence Livermore National Lab., CA (United States)

1998-08-01

Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LH{sub 2}) or ambient-temperature compressed hydrogen (CH{sub 2}). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (lower energy requirement for hydrogen liquefaction and reduced evaporative losses). This paper shows an evaluation of the applicability of the insulated pressure vessels for light-duty vehicles. The paper shows an evaluation of evaporative losses and insulation requirements and a description of the current analysis and experimental plans for testing insulated pressure vessels. The results show significant advantages to the use of insulated pressure vessels for light-duty vehicles.

Earthquake-proof supporting structure in reactor vessel

International Nuclear Information System (INIS)

Sakurai, Akio; Sekine, Katsuhisa; Madokoro, Manabu; Katoono, Shin-ichi; Konno, Mutsuo; Suzuki, Takuro.

1990-01-01

Conventional earthquake-proof structure comprises a vessel vibration stopper integrated to a reactor vessel, powder for restricting the horizontal displacements, a safety vessel surrounds the outer periphery of the reactor vessel and a safety vessel vibration stopper integrated therewith, which are fixed to buildings. However, there was a problem that a great amount of stresses are generated in the base of the reactor vessel vibration stopper due to reaction of the powders which restrict thermal expansion. In order to remarkably reduce the reaction of the powers, powders are charged into a spaces formed between each of the reactor vessel vibration stopper, the safety vessel vibration stopper and the flexible member disposed between them. According to this constitution, the reactor vessel vibration stopper does not undergo a great reaction of the powers upon thermal expansion of the reactor vessel to moderate the generated stresses, maintain the strength and provide earthquake-proof supporting function. (N.H.)

Ex-vessel corium coolability sensitivity study with the CORQUENCH code

International Nuclear Information System (INIS)

Robb, Kevin; Corradini, Michael

2009-01-01

An unresolved safety issue for light water reactor beyond design basis accidents is the coolability and stabilization of ex-vessel core melt debris by top flooding. Several experimental programs, including the OECD MACE, MCCI-1, and the current MCCI-2 program, have investigated core-concrete interactions and debris cooling of ex-vessel core melts. As part of the OECD programs, the CORQUENCH computer model was developed based on phenomena identified from the experiments. Predictions by CORQUENCH have previously been compared against experiments and have also been extrapolated to reactor scale. The current study applied statistical techniques to investigate the importance of initial system parameters and cooling phenomena in CORQUENCH 3.01 on the accident progression of ex-vessel core melts. The purpose of this sensitivity study is to identify parameters that are of major importance, any code peculiarities over the range of inputs, and where modeling improvements may produce the most gain in prediction accuracy. The sensitivity studies were carried out over a range of input conditions, in 1-D and 2-D geometries, and for two concrete compositions. In terms of initial system parameters, the melt height had the most importance on concrete ablation and melt coolability. With respect to cooling phenomena, the amount of melt entrainment through the crust had the most importance on concrete ablation and melt coolability. (author)

Pressure vessel for nuclear reactors

International Nuclear Information System (INIS)

1975-01-01

The invention applies to a pressure vessel for nuclear reactors whose shell, made of cast metal segments, has a steel liner. This liner must be constructed to withstand all operational stresses and to be easily repairable. The invention solves this problem by installing the liner at a certain distance from the inner wall of the pressure vessel shell and by filling this clearance with supporting concrete. Both the concrete and the steel liner must have a lower prestress than the pressure vessel shell. In order to avoid damage to the liner when prestressing the pressure vessel shell, special connecting elements are provided which consist of welded-on fastening elements projecting into recesses in the cast metal segments of the pressure vessel. Their design is described in detail. (TK) [de

Safety vessels for explosive fusion reactor

International Nuclear Information System (INIS)

Mineev, V.

1994-01-01

The failure of several types of geometrically similar cylindrical and spherical steel and glass fibers vessels filled with water or air was investigated when an explosive charge of TNT was detonated in the center. Vessels had radius 50-1000 mm, thickness of walls 2-20%. The detonation on TNT imitated energy release. The parameter: K = M/mf is a measure of the strength of the vessel where M is the mass of the vessel, and mf is the mass of TNT for which the vessel fails. This demanded 2-4 destroyed and nondestroyed shots. It may be showed that: K=A/σ f where σ f is the fracture stress of the material vessel, and A = const = F(energy TNT, characteristic of elasticity of vessel material). The chief results are the following: (1) A similar increase in the geometrical dimensions of steel vessels by a factor of 10 leads to the increase of parameter K in about 5 times and to decrease of failure deformation in 7 times (scale effect). (2) For glass fibers, scale effect is absent. (3) This problem is solved in terms of theory energetic scale effect. (4) The concept of TNT equivalent explosive makes it possible to use these investigations to evaluate the response of safety vessels for explosive fusion reactor

Experience with the WWER-440 MW reactor pressure vessel in-service inspections and evaluation of their results

International Nuclear Information System (INIS)

Brumovsky, M.; Kralovec, J.; Prepechal, J.; Sulc, J.

1989-01-01

The Power Machinery Plant of Skoda Works in Plzen carries out in-service inspections of WWER-440 MW reactor pressure vessels by means of remote controlled inspection equipment - the TRC reactor test system, and some other inspections devices. The results of the in-service inspections were evaluated by methods based on the fracture mechanics approach, the knowledge of stress and strain distribution, and the operating history of the pressure vessels. Examples of types of defects found and their analysis are shown. (author). 1 tab

An experimental study on in-vessel debris retention through gap cooling

International Nuclear Information System (INIS)

Kang, Kyung Ho; Kim, Jong Whan; Cho, Young Ro; Chang, Young Cho; Park, Rae Jun; Gu, Kil Mo; Kim, Sang Baik; Kim, Hee Dong

1999-04-01

LAVA experiments have been performed using high temperature molten material to be relocated into the 1/8 linear scaled vessel of a reactor lower plenum filled with water. An Al 2 O 3 /Fe tehrmite melt (Al 2 O 3 only) was used as a corium simulant. In this study, the influence of various initial conditions, such as internal pressure load across the vessel wall, the material composition of the melt simulant, water subcooling and depth on gap formation were investigated. As well, the thermal and mechanical behaviors of the vessel were examined. In case the internal pressure load was imposed, the gap formation between the continuous solidified debris and the vessel wall was clearly shown with post-test examination. On the other hand, in case the internal pressure load was not imposed, the iron welded to the inner surface of the vessel and the vessel experienced ablation to about 5 mm. The cooling rate of the vessel was very slow in the tests using Al 2 O 3 /Fe thermite melt but it was rather fast using Al 2 O 3 melt. It is postulated that in the Al 2 O 3 /Fe thermite tests, the iron melt layer is so dense that water ingression into the gap is difficult due to the high pressure of escaping steam. On the other hand, in the porosity of an Al 2 O 3 melt layer could enhance water ingression into the gap by giving the flow path of the evaporated steam through the porous media. The water height and subcooling could affect the melt pool formation and the initial thermal attack to the vessel. However, at present stage, the effect of water subcooling on the thermal behavior of the vessel couldn't be generalized. For clear confirmation of the effect of water subcooling, the tests will be performed at the saturated water condition. (author). 19 refs., 3 tabs., 36 figs

Ultrasonic testing of electron beam closure weld on pressure vessel

International Nuclear Information System (INIS)

Andrews, R.W.

1975-01-01

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

Investigation and analysis on ITER in-vessel coils’ raw-materials

International Nuclear Information System (INIS)

Jin, Huan; Wu, Yu; Long, Feng; Yu, Min; Han, Qiyang; Liu, Huajun

2013-01-01

Highlights: • The R and D works for the ITER in-vessel coils (IVC) are now being conducted in Institute of Plasma Physics, and the analysis work are being done by Princeton Plasma Physics Laboratory. • There is little published paper about the raw materials for ITER IVC coils. • This manuscript points out the progress of the selected materials for ITER IVC coils. -- Abstract: The ITER in-vessel coils (IVCs) consist of 27 coils edge localized modes (ELM) and 2 coils vertical stabilization (VS) which are all mounted on the vacuum vessel wall behind the shield modules. The IVCs design and manufacturing work is being conducted in between Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and Princeton Plasma Physics Laboratory (PPPL). Because the position of ELM and VS coils is close and face to the plasma, the IVCs must undergo a severe environment, such as the high dose of radiation and high operation temperature, thus the conventional electrical insulation materials cannot be used. And the technology of “Stainless Steel Jacketed Mineral Insulated Conductor” (SSMIC) is deemed as the best choice to provide the necessary radiation resistance and compatibility strength in ITER's vacuum vessel. While mineral insulated conductor technology is not new, and is similar to the mineral insulated cable used in industrial. Some difficulties still need to be solved, such as searching for the proper raw-materials to make sure that the conductor have the properties of high current carrying capability, the necessary radiation resistance, the proper strength, at the same time, it must be come true in manufacture technology. This paper described the analysis of the materials for VS and ELM coil conductor

Multiphase flow in ex-vessel coolability: development of an innovative concept

International Nuclear Information System (INIS)

Corradini, Michael L.

2006-01-01

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific Advanced Light Water Reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The paper provides the background of past experiments as well as key fundamentals that are needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability

Clinical results of single-vessel versus multiple-vessel infrapopliteal intervention

OpenAIRE

Darling, Jeremy; McCallum, John C.; Soden, Peter A.; Hon, J.J. (John J.); Guzman, R.J. (Raul J.); Wyers, M.C. (Mark C.); Verhagen, Hence; Schermerhorn, Marc

2016-01-01

textabstractObjective The effects of concomitant endovascular interventions on multiple infrapopliteal vessels are not well known, and the short-term and long-term sequelae of such procedures have not been reported. Methods From 2004 to 2014, 673 limbs in 528 patients underwent an infrapopliteal endovascular intervention for tissue loss (77%), rest pain (13%), stenosis of a previously treated vessel (5%), acute limb ischemia (3%), or claudication (2%). Outcomes included wound healing, RAS eve...

Development of LILAC-meltpool for the thermo-hydraulic analysis of core melt relocated in a reactor vessel

International Nuclear Information System (INIS)

Kim, Jong Tae; Kim, Sang Baik; Kim, Hee Dong

2002-03-01

LILAC-meltpool has been developed to study thermo-hydraulic behavior of molten pool and thermal behavior of vessel wall during severe accident. To validate LILAC-meltpool code several two and three dimensional thermo-hydraulic problems were selected and solved. The benchmark problems have experimental results or verified numerical results. Through the validation it was found that LILAC-meltpool reproduces very accurate numerical results. Two-layered semicircular pool was solved to study thermal and hydraulic characteristics of pool stratification. The LAVA experiment using alumina/ferrite molten pool was calculated and compared with computed results. Cooling of alumina/ferrite two-layered pool was affected by stratification. In the numerical results temperature of vessel inner was highest at a location below the interface. Crust was developed from upper surface and lower outer surface, but in the area near the interface corium simulant existed as molten state for long time. LAVA-4 experiment was studied using gap-cooling model in LILAC-meltpool code. Temperature increase of LAVA vessel after alumina melt relocation was strongly dependent on gap formation mechanism. Calculated cooling rates of the vessel were very similar to experimental results. For LAVA experiments which do not have heat generation coolant penetrates easily into a gap and it is found that gap-cooling is very effective for cooling of vessel, but it is thought that coolant penetration could be limited near upper part of gap because of decay heat and high temperature of corium crust

The Use of Prestressed Concrete Vessels in the French Power Reactor Programme

International Nuclear Information System (INIS)

Conte, F.; Dambrine, C.; Gaussot, D.

1963-01-01

This paper deals with the use of pre-stressed concrete for the G2 and G3 reactors at Marcoule and for the EDF3 reactor now under construction at Chinon. The first two reactors have been operating at power since 1959 and 1960 respectively. Messrs. Conte and Dambrine discuss the problems that arose during construction of the vessels for G2 and G3 and also deal with the experience gained in operation - experience which suggests that they are extremely safe- Work on the EDF3 vessel, begun at Chinon in the second half of 1961, is still under way and should be finished towards the end of 1963. Mr. Gaussot discusses the reasons for choosing this type of vessel, the results of calculations and mock-up tests, and the problems presented by the construction itself. A number of studies have been devoted to the future prospects of prestressed concrete structures for reactors. It would seem that working pressures could be increased, if desired, and, in any case, that dimensions could be considerably enlarged, thus offering the chance of integral-type solutions. (author) [fr

Hydraulic nuts (HydraNuts) for reactor vessel tensioning

International Nuclear Information System (INIS)

Greenwell, Steve

2008-01-01

The paper will present how the introduction of hydraulic nuts - HydraNuts, has reduced critical path times, dose exposure for workers and improved working safety conditions around the reactor vessel during tensioning or de-tensioning operations. It will focus upon detailing the advantages realized by utilities that have introduced the technology and providing examples of the improvements made to the process as well as discussing the engineering design change packages required to make the conversion to the new system. HydraNuts replace the traditional mechanical nut/stud tensioning equipment, combining the two functions into a single system, designed for easy installation and operation by one individual. The primary components of the HydraNut can be assembled without the need for external crane or hoist support and are designed so that each sub assembly can be fitted separately. Once all HydraNuts are fitted to the Rx vessel studs and are sitting on the main Rx vessel head flange, then a system of flexible hydraulic hoses is connected to them, forming a closed loop hydraulic harness, which will allow for simultaneous pressurization of all HydraNuts. Hydraulic pressure is obtained by the use of a hydraulic pumping unit and the resultant load generated in each HydraNut is transferred to the stud and main flange closure is obtained. While maintaining hydraulic pressure, a locking ring is rotated into place on the HydraNut assembly that will support the tensioned load mechanically when the hydraulic pressure is released from the hose harness assembly. The hose harness is removed and the HydraNut is now functioning as a mechanical nut retaining the tensioned load. The HydraNut system for Rx vessel applications was first introduced into a plant in the U.S. in October 2006 and based upon the benefits realized subsequent projects are under way within the Asian and U.S. operating fleet. (author)

Vessel size effect on the characteristic frequency of the free surface fluctuations

International Nuclear Information System (INIS)

Nam, Ho Yun; Kim, Min Joon; Kim, Jong Man; Choi, Byoung Hae

2004-01-01

Studies of the free surface fluctuations is one of the important topics in a liquid metal nuclear reactor using sodium as the coolant that has a free surface in the upper plenum of the reactor vessel. The main reasons for the study on the free surface fluctuations can be summarized as: 1. to secure the structural integrity of a reactor vessel by considering the thermal stress on the vessel wall induced by the fluctuations of the free surface between the hot sodium and cold cover gas, 2. to prevent the cover gas entrainment at the free surface of the sodium because the entrained gas causes a change in the reactivity and also reduces the heat removal capability in the core. Some experimental studies on the free surface fluctuations have been reported. However, most of them focus on the gas entrainment phenomena and only a few works concern the basic characteristics of the free surface fluctuations. Since the thermal stress on the wall is strongly dependent on the amplitude and frequency of the free surface fluctuations, studies on the amplitudes and frequencies should receive more attention. In Nam, empirical formulae on the amplitudes and frequencies with respect to the geometric and hydraulic parameters were introduced. It is an interesting result, but the experiment was performed within the parameter range near the onset point of the fluctuations. In the real reactor condition, larger sized fluctuations may exist and the formula needs to be modified. In this study, we performed experiments on the free surface fluctuations, especially on larger sized fluctuations and made an analysis of the amplitudes and frequencies. The main focus of this paper is the effect of the vessel size on the characteristic frequencies. It is thought to be helpful for finding the scaling laws, for example, designing a scale-down experiment

Americium behaviour in plastic vessels.

Science.gov (United States)

Legarda, F; Herranz, M; Idoeta, R; Abelairas, A

2010-01-01

The adsorption of (241)Am dissolved in water in different plastic storage vessels was determined. Three different plastics were investigated with natural and distilled waters and the retention of (241)Am by these plastics was studied. The same was done by varying vessel agitation time, vessel agitation speed, surface/volume ratio of water in the vessels and water pH. Adsorptions were measured to be between 0% and 70%. The adsorption of (241)Am is minimized with no water agitation, with PET or PVC plastics, and by water acidification. Copyright 2009 Elsevier Ltd. All rights reserved.

Neutron irradiation effects in reactor pressure vessel steels and weldments. Working document

International Nuclear Information System (INIS)

1998-10-01

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. A separate abstract was prepared for the introduction and for each of the eleven chapters, which are: 1. Reactor Pressure Vessel Design, 2. Reactor Pressure Materials, 3. WWER Pressure Vessels, 4. Determination of Mechanical Properties, 5. Neutron Exposure, 6. Methodology of Irradiation Experiments, 7. Effect of Irradiation on Mechanical Properties, 8. Mechanisms of Irradiation Embrittlement, 9. Modelling of Irradiation Damage, 10. Annealing of Irradiation Damage, 11. Safety Assessment using Surveillance Programmes and Data Bases

Vessel Sampling and Blood Flow Velocity Distribution With Vessel Diameter for Characterizing the Human Bulbar Conjunctival Microvasculature.

Science.gov (United States)

Wang, Liang; Yuan, Jin; Jiang, Hong; Yan, Wentao; Cintrón-Colón, Hector R; Perez, Victor L; DeBuc, Delia C; Feuer, William J; Wang, Jianhua

2016-03-01

This study determined (1) how many vessels (i.e., the vessel sampling) are needed to reliably characterize the bulbar conjunctival microvasculature and (2) if characteristic information can be obtained from the distribution histogram of the blood flow velocity and vessel diameter. Functional slitlamp biomicroscope was used to image hundreds of venules per subject. The bulbar conjunctiva in five healthy human subjects was imaged on six different locations in the temporal bulbar conjunctiva. The histograms of the diameter and velocity were plotted to examine whether the distribution was normal. Standard errors were calculated from the standard deviation and vessel sample size. The ratio of the standard error of the mean over the population mean was used to determine the sample size cutoff. The velocity was plotted as a function of the vessel diameter to display the distribution of the diameter and velocity. The results showed that the sampling size was approximately 15 vessels, which generated a standard error equivalent to 15% of the population mean from the total vessel population. The distributions of the diameter and velocity were not only unimodal, but also somewhat positively skewed and not normal. The blood flow velocity was related to the vessel diameter (r=0.23, Psampling size of the vessels and the distribution histogram of the blood flow velocity and vessel diameter, which may lead to a better understanding of the human microvascular system of the bulbar conjunctiva.

Improving methodology in open vessel digestion with a graphite heating block (T7)

International Nuclear Information System (INIS)

Kainrath, P.; Conrads, B.; Ross, A.

2002-01-01

Full text: Open block digestion systems have been very popular in environmental analysis over the past decades, but have consistently suffered from the major drawback of their sensitivity against corrosion and the subsequent risk of contamination. Therefore block digestion systems have not been considered state-of-the-art technology in trace and ultra trace sample preparation. Graphite block digestion systems are well established in North America and are recently becoming more frequently considered in Europe. These systems overcome the deficiencies of the traditional systems, made from stainless steel or aluminum, because the block is manufactured from graphite and typically coated with a fluoro-polymer to present the possibility metallic contamination from the surface of the system during the handling of the samples. Graphite block systems present an alternative to the current mainstream technology of open and closed vessel microwave assisted digestion systems, as they allow large numbers of samples to be digested simultaneously, thus overcoming one of the major weaknesses of closed vessel systems. More recently a number of improvements in the technology has been developed for graphite block digestion systems and studies have been performed to evaluate the effects of such improvements. The paper presented will deal with the technological improvements: monitoring and control of sample temperature vs. monitoring of block temperature, elimination of cross contamination effects during open vessel block digestion, evaporation of samples for pre-concentration or multiple digestion steps, addressing the needs of various labs and applications for block digesters. The effects of those developments will be discussed; application examples and finally an outlook into possible future trends for graphite block digestion systems will be given. (author)

High-Speed Vessel Noises in West Hong Kong Waters and Their Contributions Relative to Indo-Pacific Humpback Dolphins (Sousa chinensis

Directory of Open Access Journals (Sweden)

Paul Q. Sims

2012-01-01

Full Text Available The waters of West Hong Kong are home to a population of Indo-Pacific humpback dolphins (Sousa chinensis that use a variety of sounds to communicate. This area is also dominated by intense vessel traffic that is believed to be behaviorally and acoustically disruptive to dolphins. While behavioral changes have been documented, acoustic disturbance has yet to be shown. We compared the relative sound contributions of various high-speed vessels to nearby ambient noise and dolphin social sounds. Ambient noise levels were also compared between areas of high and low traffic. We found large differences in sound pressure levels between high traffic and no traffic areas, suggesting that vessels are the main contributors to these discrepancies. Vessel sounds were well within the audible range of dolphins, with sounds from 315–45,000 Hz. Additionally, vessel sounds at distances ≥100 m exceeded those of dolphin sounds at closer distances. Our results reaffirm earlier studies that vessels have large sound contributions to dolphin habitats, and we suspect that they may be inducing masking effects of dolphin sounds at close distances. Further research on dolphin behavior and acoustics in relation to vessels is needed to clarify impacts.

Radiation embrittlement of PWR vessel supports

International Nuclear Information System (INIS)

Cheverton, R.D.; Robinson, G.C.; Pennell, W.E.; Nanstad, R.K.

1989-01-01

Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100 degree C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Probabilistic approach to the analysis of reactor pressure vessel integrity during a pressurized thermal shock

International Nuclear Information System (INIS)

Adamec, P.

2000-12-01

Following a general summary of the issue, an overview of international experience (USA; Belgium, France, Germany, Russia, Spain, Sweden, The Netherlands, and the UK; and probabilistic PTS assessment for the reactor pressure vessel at Loviisa-1, Finland) is presented, and the applicable computer codes (VISA-II, OCA-P, FAVOR, ZERBERUS) are highlighted and their applicability to VVER type reactor pressure vessels is outlined. (P.A.)

Supervised retinal vessel segmentation from color fundus images based on matched filtering and AdaBoost classifier.

Directory of Open Access Journals (Sweden)

Nogol Memari

Full Text Available The structure and appearance of the blood vessel network in retinal fundus images is an essential part of diagnosing various problems associated with the eyes, such as diabetes and hypertension. In this paper, an automatic retinal vessel segmentation method utilizing matched filter techniques coupled with an AdaBoost classifier is proposed. The fundus image is enhanced using morphological operations, the contrast is increased using contrast limited adaptive histogram equalization (CLAHE method and the inhomogeneity is corrected using Retinex approach. Then, the blood vessels are enhanced using a combination of B-COSFIRE and Frangi matched filters. From this preprocessed image, different statistical features are computed on a pixel-wise basis and used in an AdaBoost classifier to extract the blood vessel network inside the image. Finally, the segmented images are postprocessed to remove the misclassified pixels and regions. The proposed method was validated using publicly accessible Digital Retinal Images for Vessel Extraction (DRIVE, Structured Analysis of the Retina (STARE and Child Heart and Health Study in England (CHASE_DB1 datasets commonly used for determining the accuracy of retinal vessel segmentation methods. The accuracy of the proposed segmentation method was comparable to other state of the art methods while being very close to the manual segmentation provided by the second human observer with an average accuracy of 0.972, 0.951 and 0.948 in DRIVE, STARE and CHASE_DB1 datasets, respectively.

46 CFR 4.03-40 - Public vessels.

Science.gov (United States)

2010-10-01

... INVESTIGATIONS Definitions § 4.03-40 Public vessels. Public vessel means a vessel that— (a) Is owned, or demise... Department (except a vessel operated by the Coast Guard or Saint Lawrence Seaway Development Corporation...

Vessel bifurcation localization based on intraoperative three-dimensional ultrasound and catheter path for image-guided catheter intervention of oral cancers.

Science.gov (United States)

Luan, Kuan; Ohya, Takashi; Liao, Hongen; Kobayashi, Etsuko; Sakuma, Ichiro

2013-03-01

We present a method to localize intraoperative target vessel bifurcations under bones for ultrasound (US) image-guided catheter interventions. A catheter path is recorded to acquire skeletons for the target vessel bifurcations that cannot be imaged by intraoperative US. The catheter path is combined with the centerlines of the three-dimensional (3D) US image to construct a preliminary skeleton. Based on the preliminary skeleton, the orientations of target vessels are determined by registration with the preoperative image and the bifurcations were localized by computing the vessel length. An accurate intraoperative vessel skeleton is obtained for correcting the preoperative image to compensate for vessel deformation. A reality check of the proposed method was performed in a phantom experiment. Reasonable results were obtained. The in vivo experiment verified the clinical workflow of the proposed method in an in vivo environment. The accuracy of the centerline length of the vessel for localizing the target artery bifurcation was 2.4mm. These results suggest that the proposed method can allow the catheter tip to stop at the target artery bifurcations and enter into the target arteries. This method can be applied for virtual reality-enhanced image-guided catheter intervention of oral cancers. Copyright © 2013 Elsevier Ltd. All rights reserved.

DOMPAC dosimetry experiment. Neutronic simulation of the thickness of a PWR pressure vessel. Irradiation damages

International Nuclear Information System (INIS)

Alberman, A.; Faure, M.; Thierry, M.; Hoclet, O.; Le Dieu de Ville, A.; Nimal, J.C.; Soulat, P.

1979-01-01

For suitable extrapolation of irradiated PWR ferritic steel results, proper irradiation of the pressure vessel has been 'simulated' in test reactor. For this purpose, a huge steel block (20 cm in depth) was loaded with Saclay's graphite (GAMIN) and tungsten damage detectors. Core-block water gap was optimized through spectrum indexes method, by ANISN and SABINE codes so that spectrum in 1/4 thickness matches with ANISN computations for PWR Fessenheim 1. A good experimental agreement is found with calculated dpa damage gradient. 3D Monte Carlo computation (TRIPOLI), was performed on the DOMPAC device, and spectrum indexes evolution was found consistent with experimental results. Surveillance rigs behind a 'thermal shield' were also simulated, including damage and activation monitors. Dosimetry results give an order of magnitude of accuracies involved in projecting steel sample embrittlement to the pressure vessel [fr

Prosopomorphic vessels from Moesia Superior

Directory of Open Access Journals (Sweden)

Nikolić Snežana

2008-01-01

Full Text Available The prosopomorphic vessels from Moesia Superior had the form of beakers varying in outline but similar in size. They were wheel-thrown, mould-made or manufactured by using a combination of wheel-throwing and mould-made appliqués. Given that face vessels are considerably scarcer than other kinds of pottery, more than fifty finds from Moesia Superior make an enviable collection. In this and other provinces face vessels have been recovered from military camps, civilian settlements and necropolises, which suggests that they served more than one purpose. It is generally accepted that the faces-masks gave a protective role to the vessels, be it to protect the deceased or the family, their house and possessions. More than forty of all known finds from Moesia Superior come from Viminacium, a half of that number from necropolises. Although tangible evidence is lacking, there must have been several local workshops producing face vessels. The number and technological characteristics of the discovered vessels suggest that one of the workshops is likely to have been at Viminacium, an important pottery-making centre in the second and third centuries.

Recovery process of wall condition in KSTAR vacuum vessel after temporal machine-vent for repair

Energy Technology Data Exchange (ETDEWEB)

Kim, Kwang Pyo, E-mail: kpkim@nfri.er.ke; Hong, Suk-Ho; Lee, Hyunmyung; Song, Jae-in; Jung, Nam-Yong; Lee, Kunsu; Chu, Yong; Kim, Hakkun; Park, Kaprai; Oh, Yeong-Kook

2015-10-15

Highlights: • Efforts have been made to obtain vacuum condition that is essential for the plasma experiments. • For example, the vacuum vessel should be vented to repair in-vessel components such as diagnostic shutter, and PFC damaged by high energy plasma. • Here, we present the recovery process of wall condition in KSTAR after temporal machine-vent for repair. • It is found that an acceptable vacuum condition has been achieved only by plasma based wall conditioning techniques such as baking, GDC, and boronization. • This study was that the proper recovering method of the vacuum condition should be developed according to the severity of the accident. - Abstract: Efforts have been made to obtain vacuum condition that is essential for the plasma experiments. Under certain situations, for example, the vacuum vessel should be vented to repair in-vessel components such as diagnostic shutter, exchange of window for diagnostic equipment, and PFC damaged by high energy plasma. For the quick restart of the campaign, a recovery process was established to make the vacuum condition acceptable for the plasma experiment. In this paper, we present the recovery process of wall condition in KSTAR after temporal machine-vent for repair. It is found that an acceptable vacuum condition has been achieved only by plasma based wall conditioning techniques such as baking, GDC, and boronization. This study was that the proper recovering method of the vacuum condition should be developed according to the severity of the accident.

Dalton's disputed nitric oxide experiments and the origins of his atomic theory.

Science.gov (United States)

Usselman, Melvyn C; Leaist, Derek G; Watson, Katherine D

2008-01-11

In 1808 John Dalton published his first general account of chemical atomic theory, a cornerstone of modern chemistry. The theory originated in his earlier studies of the properties of atmospheric gases. In 1803 Dalton discovered that oxygen combined with either one or two volumes of nitric oxide in closed vessels over water and this pioneering observation of integral multiple proportions provided important experimental evidence for his incipient atomic ideas. Previous attempts to reproduce Dalton's experiments have been unsuccessful and some commentators have concluded the results were fraudulent. We report a successful reconstruction of Dalton's experiments and provide an analysis exonerating him of any scientific misconduct. But we conclude that Dalton, already thinking atomistically, adjusted experimental conditions to obtain the integral combining proportions.

Nuclear reactor pressure vessel flaw distribution development

International Nuclear Information System (INIS)

Kennedy, E.L.; Foulds, J.R.; Basin, S.L.

1991-12-01

Previous attempts to develop flaw distributions for probabilistic fracture mechanics analyses of pressurized water reactor (PWR) vessels have aimed at the estimation of a ''generic'' distribution applicable to all PWR vessels. In contrast, this report describes (1) a new flaw distribution development analytic methodology that can be applied to the analysis of vessel-specific inservice inspection (ISI) data, and (2) results of the application of the methodology to the analysis of flaw data for each vessel case (ISI data on three PWR vessels and laboratory inspection data on sections of the Midland reactor vessel). Results of this study show significant variation among the flaw distributions derived from the various data sets analyzed, strongly suggesting than a vessel-specific flaw distribution (for vessel integrity prediction under pressurized thermal shock) is preferred over a ''generic'' distribution. In addition, quantitative inspection system flaw sizing accuracy requirements have been identified for developing a flaw distribution from vessel ISI data. The new flaw data analysis methodology also permits quantifying the reliability of the flaw distribution estimate. Included in the report are identified needs for further development of several aspects of ISI data acquisition and vessel integrity prediction practice

A tritium vessel cleanup experiment in TFTR

International Nuclear Information System (INIS)

Caorlin, M.; Kamperschroer, J.; Owens, D.K.; Voorhees, D.; Mueller, D.; Ramsey, A.T.; La Marche, P.H.; Loughlin, M.J.

1995-03-01

A simple tritium cleanup experiment was carried out in TFTR following the initial high power deuterium-tritium discharges in December 1993. A series of 34 ohmic and deuterium neutral beam fueled shots was used to study the removal of tritium implanted into the wall and limiters. A very large plasma was created in each discharge to ''scrub'' an area as large as possible. Beam-fueled shots at 2.5 to 7.5 MW of injected power were used to monitor tritium concentration levels in the plasma by detection of DT-neutrons. The neutron signal decreased by a factor of 4 during the experiment, remaining well above the expected T-burnup level. The amount of tritium recovered at the end of the cleanup was about 8% of the amount previously injected with high power DT discharges. The experience gained suggests that measurements of tritium inventory in the torus are very difficult to execute and require dedicated systems with overall accuracy of 1%

Targeting Therapy Resistant Tumor Vessels

Science.gov (United States)

2008-08-01

Morris LS. Hysterectomy vs. resectoscopic endometrial ablation for the control of abnormal uterine bleeding . A cost-comparative study. J Reprod Med 1994;39...after the antibody treatment contain a pericyte coat, vessel architecture is normal, the diameter of the vessels is smaller (dilated, abnormal vessels...involvement of proteases from inflammatory mast cells and functionally abnormal (Carmeliet and Jain, 2000; Pasqualini (Coussens et al., 1999) and other bone

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)

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)

Flaw distribution development from vessel ISI data

International Nuclear Information System (INIS)

Foulds, J.R.; Kennedy, E.L.; Basin, S.L.; Rosinski, S.T.

1991-01-01

Previous attempts to develop flaw distributions for use in the structural integrity evaluation of pressurized water reactor (PWR) vessels have aimed at the estimation of a ''generic'' distribution applicable to all vessels. In contrast, this paper describes the analysis of vessel-specific in-service inspection (ISI) data for the development of a flaw distribution reliably representative of the condition of the particular vessel inspected. The application of the methodology may be extended to other vessels, but has been primarily developed for PWR reactor vessels. For this study, the flaw data analyzed included data obtained from three recently performed PWR vessel ISIs and from laboratory inspection of selected weldment sections of the Midland reactor vessel. The variability in both the character of the reviewed data (size range of flaws, number of flaws) and the UT (ultrasonic test) inspection system performance identified a need for analyzing the inspection results on a vessel-, or data set-specific basis. For this purpose, traditional histogram-based methods were inadequate, and a new methodology that can accept a very small number of flaws (typical of vessel-specific ISI results) and that includes consideration of inspection system flaw detection reliability, flaw sizing accuracy and flaw detection threshold, was developed. Results of the application of the methodology to each of the four PWR reactor vessel cases studied are presented and discussed

Seals for sealing a pressure vessel such as a nuclear reactor vessel or the like

International Nuclear Information System (INIS)

Bruns, H.J.; Huelsermann, K.H.

1975-01-01

A description is given of seals for sealing a pressure vessel such as a nuclear reactor vessel, steam boiler vessel, or any other vessel which is desirably sealed against pressure of the type including a housing and a housing closure that present opposed vertical sealing surfaces which define the sides of a channel. The seals of the present invention comprise at least one sealing member disposed in the channel, having at least one stop face, a base portion and two shank portions extending from the base portion to form a groove-like recess. The shank portions are provided with sealing surfaces arranged to mate with the opposed vertical pressure vessel sealing surfaces. A shank-spreading wedge element also disposed in the channel has at least one stop face and is engaged in the groove-like recess with the sealing member and wedge element stop face adjacent to each other

Proposal of Ex-Vessel dosimetry for pressure vessel Atucha II

International Nuclear Information System (INIS)

Chiaraviglio, N.; Bazzana, S.

2013-01-01

Nuclear reactor dosimetry has the purpose of guarantee that changes in material mechanical properties of critical materials do not compromise the reactor safety. In PWR in which the top of the reactor vessel is open once a year, is possible to use Charpy specimens to measure the change in mechanical properties. Atucha II nuclear power plant is a reactor with on-line refueling so there is no access to the inside of the pressure vessel. Because of this, ex-vessel dosimetry must be performed and mechanical properties changes must be inferred from radiation damage estimations. This damage can be calculated using displacement per atom cross sections and a transport code such as MCNP. To increase results reliability it is proposed to make a neutron spectrum unfolding using activation dosimeters irradiated during one operation cycle of the power plant. In this work we present a dosimetry proposal for such end, made in base of unfolding procedures and experimental background. (author) [es

Forging technology for large nuclear pressure vessel parts

International Nuclear Information System (INIS)

Kakimoto, Hideki; Ikegami, Tomonori

2014-01-01

The increasing output of nuclear power generation calls for larger vessels for next-generation nuclear power plants. A vessel with an increased diameter requires increased load for its forging, which can make it difficult to use a conventional solid die. In order to reduce the forging load, a rotary incremental forging method has been applied to hot forging. This method includes pressing and rotating a material in an incremental manner such that a target shape is obtained. This study aimed at improving the accuracy of numerical simulation for the rotary incremental forging to reduce the load when forging large vessels. This has enabled the temperature of the material and flow stress to be precisely predicted; an example of this is reported in the paper. Specifically, the heat transfer coefficient to be used for the numerical simulation had been determined experimentally from a small-scale hot-forging. The reduction of the flow stress associated with incremental forging, had been deduced from a compression test, and the value was applied to the numerical simulation. A preform was designed on the basis of the above simulation to perform a 1/1 size scale experiment. A precision of better than 5% has been confirmed for the shape prediction. (author)

Acid digestion of geological and environmental samples using open-vessel focused microwave digestion.

Science.gov (United States)

Taylor, Vivien F; Toms, Andrew; Longerich, Henry P

2002-01-01

The application of open vessel focused microwave acid digestion is described for the preparation of geological and environmental samples for analysis using inductively coupled plasma-mass spectrometry (ICP-MS). The method is compared to conventional closed-vessel high pressure methods which are limited in the use of HF to break down silicates. Open-vessel acid digestion more conveniently enables the use of HF to remove Si from geological and plant samples as volatile SiF4, as well as evaporation-to-dryness and sequential acid addition during the procedure. Rock reference materials (G-2 granite, MRG-1 gabbros, SY-2 syenite, JA-1 andesite, and JB-2 and SRM-688 basalts) and plant reference materials (BCR and IAEA lichens, peach leaves, apple leaves, Durham wheat flour, and pine needles) were digested with results comparable to conventional hotplate digestion. The microwave digestion method gave poor results for granitic samples containing refractory minerals, however fusion was the preferred method of preparation for these samples. Sample preparation time was reduced from several days, using conventional hotplate digestion method, to one hour per sample using our microwave method.

High Performance Marine Vessels

CERN Document Server

Yun, Liang

2012-01-01

High Performance Marine Vessels (HPMVs) range from the Fast Ferries to the latest high speed Navy Craft, including competition power boats and hydroplanes, hydrofoils, hovercraft, catamarans and other multi-hull craft. High Performance Marine Vessels covers the main concepts of HPMVs and discusses historical background, design features, services that have been successful and not so successful, and some sample data of the range of HPMVs to date. Included is a comparison of all HPMVs craft and the differences between them and descriptions of performance (hydrodynamics and aerodynamics). Readers will find a comprehensive overview of the design, development and building of HPMVs. In summary, this book: Focuses on technology at the aero-marine interface Covers the full range of high performance marine vessel concepts Explains the historical development of various HPMVs Discusses ferries, racing and pleasure craft, as well as utility and military missions High Performance Marine Vessels is an ideal book for student...

Reactor-vessel-sectioning demonstration

International Nuclear Information System (INIS)

Lundgren, R.A.

1981-07-01

A successful technical demonstration of simulated reactor vessel sectioning was completed using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel: an air arc gouger was chosen to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. After the simulated vessel was successfully cut from the SS side, another cut was made, starting from the carbon steel side. This cut was also successful. Cutting from the carbon steel side has the advantages of cost reduction since the air arc gouging step is eliminated and contamination controlled because the molten metal is blown inward

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)

The vessel fluence; Fluence cuve

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

This book presents the proceedings of the technical meeting on the reactors vessels fluence. They are grouped in eight sessions: the industrial context and the stakes of the vessels control; the organization and the methodology for the fluence computation; the concerned physical properties; the reference computation methods; the fluence monitoring in an industrial context; vessels monitoring under irradiation; others methods in the world; the research and development programs. (A.L.B.)

Principles of Vessel Route Planning in Ice on the Northern Sea Route

Directory of Open Access Journals (Sweden)

Tadeusz Pastusiak

2016-12-01

Full Text Available A complex of ice cover characteristics and the season of the year were considered in relation to vessel route planning in ice-covered areas on the NSR. The criteria for navigation in ice - both year-round and seasonal were analyzed. The analysis of the experts knowledge, dissipated in the literature, allowed to identify some rules of route planning in ice-covered areas. The most important processes from the navigation point of view are the development and disintegration of ice, the formation and disintegration of fast ice and behavior of the ice massifs and polynyas. The optimal route is selected on basis of available analysis and forecast maps of ice conditions and ice class, draught and seaworthiness of the vessel. The boundary of the ice indicates areas accessible to vessels without ice class. Areas with a concentration of ice from 0 to 6/10 are used for navigation of vessels of different ice classes. Areas of concentration of ice from 7/10 up are eligible for navigation for icebreakers and vessels with a high ice class with the assistance of icebreakers. These rules were collected in the decision tree. Following such developed decision-making model the master of the vessel may take decision independently by accepting grading criteria of priorities resulting from his knowledge, experience and the circumstances of navigation. Formalized form of decision making model reduces risk of the "human factor" in the decision and thereby help improve the safety of maritime transport.

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

ITER vacuum vessel design (D201 subtask 1.3 and subtask 3). Final report

International Nuclear Information System (INIS)

1996-01-01

ITER Task No. D201, Vacuum Vessel Design (Subtask 1.3 and Subtask 3), was initiated to propose and evaluate local vacuum vessel reinforcement alternatives in proximity to the Neutral Beam, Radial Mid-Plane, Top, and Divertor Ports. These areas were reported to be highly stressed regions based on the results of preliminary stress analyses performed by the USHT (US Home Team) and the ITER Joint Central Team (JCT) at the Garching JWS (Joint Work Site). Initial design activities focused on the divertor port region which was reported to experience the highest stress intensities. Existing stress analysis models and results were reviewed with the USHT stress analysts to obtain an overall understanding of the vessel response to the various applied loads. These reviews indicated that the reported stress intensities in the divertor port region were significantly affected by the loads applied to the vessel in adjacent regions

Underwater laser beam welding technology for reactor vessel nozzles of PWRs

International Nuclear Information System (INIS)

Yoda, Masaki; Tamura, Masataka; Tamura, Masataka

2010-01-01

Toshiba has developed an underwater laser beam welding technology for the maintenance of reactor vessel nozzles of pressurized water reactors (PWRs), which eliminates the need for the drainage of water from the reactor vessel. The new welding system makes it possible to both reduce the work period and minimize the radiation exposure of workers compared with conventional technologies for welding in ambient air. We have confirmed the effectiveness of this technology through experiments in which stress corrosion cracking (SCC) was mitigated on the inner surfaces of nozzles. We are promoting its practical application in Japan and overseas in cooperation with Westinghouse Electric Company, a group company of Toshiba. (author)

Expanded Fermilab pressure vessel directory program

Energy Technology Data Exchange (ETDEWEB)

Tanner, A.

1983-01-01

Several procedures have been written to manage the information pertaining to the vacuum tanks and pressure vessels for which the laboratory is responsible. These procedures have been named TANK1 for the vessels belonging to the Accelerator Division, TANK2 and TANK3 for the vessels belonging to the Research Division and to Technical Support respectively, and TANK4 for the vessels belonging to the Business Division. The operating procedures are otherwise identical in every respect.

Expanded Fermilab pressure vessel directory program

International Nuclear Information System (INIS)

Tanner, A.

1983-01-01

Several procedures have been written to manage the information pertaining to the vacuum tanks and pressure vessels for which the laboratory is responsible. These procedures have been named TANK1 for the vessels belonging to the Accelerator Division, TANK2 and TANK3 for the vessels belonging to the Research Division and to Technical Support respectively, and TANK4 for the vessels belonging to the Business Division. The operating procedures are otherwise identical in every respect

The transition to farming and the ceramic trajectories in Western Eurasia. From ceramic figurines to vessels

Directory of Open Access Journals (Sweden)

Mihael Budja

2006-12-01

Full Text Available In Eurasia the invention of ceramic technology and production of fired-clay vessels has not necessarily been related to the dynamics of the transition to farming. The invention of ceramic technology in Europe was associated with female and animal figurine making in Gravettian technocomplex. The fired-clay vessels occurred first in hunter-gatherer contexts in Eastern Eurasia a millennia before the agriculture. The adoption of pottery making in Levant seems to correlate with the collapse of the ‘ritual economy’, social decentralisation and community fragmentation in the Levantine Pre-Pottery Neolithic. In South-eastern Europe the adoption of pottery making was closely associated with social, symbolic and ritual hunter-gatherers’ practices.

Effect of the in- and ex-vessel dual cooling on the retention of an internally heated melt pool in a hemispherical vessel

Energy Technology Data Exchange (ETDEWEB)

Ahn, K.I.; Kim, B.S.; Kim, D.H. [Korea Atomic Energy Research Inst., Thermal Hydraulic Safety Research, Taejon (Korea, Republic of)

2001-07-01

A concept of in-vessel melt retention (IVMR) by in-vessel reflooding and/or reactor cavity flooding has been considered as one of severe accident management strategies and intensive researches to be performed worldwide. This paper provides some results of analytical investigations on the effect of both in- / ex-vessel cooling on the retention of an internally heated molten pool confined in a hemispherical vessel and the related thermal behavior of the vessel wall. For the present analysis, a scale-down reactor vessel for the KSNP reactor design of 1000 MWe (a large dry PWR) is utilized for a reactor vessel. Aluminum oxide melt simulant is also utilized for a real corium pool. An internal power density in the molten pool is determined by a simple scaling analysis that equates the heat flux on the the scale-down vessel wall to that estimated from KSNP. Well-known temperature-dependent boiling heat transfer curves are applied to the in- and ex-vessel cooling boundaries and radiative heat transfer has been only considered in the case of dry in-vessel. MELTPOOL, which is a computational fluid dynamics (CFD) code developed at KAERI, is applied to obtain the time-varying heat flux distribution from a molten pool and the vessel wall temperature distributions with angular positions along the vessel wall. In order to gain further insights on the effectiveness of in- and ex-vessel dual cooling on the in-vessel corium retention, four different boundary conditions has been considered: no water inside the vessel without ex-vessel cooling, water inside the vessel without ex-vessel cooling, no water inside the vessel with ex-vessel cooling, and water inside the vessel with ex-vessel cooling. (authors)

Effect of the in- and ex-vessel dual cooling on the retention of an internally heated melt pool in a hemispherical vessel

International Nuclear Information System (INIS)

Ahn, K.I.; Kim, B.S.; Kim, D.H.

2001-01-01

A concept of in-vessel melt retention (IVMR) by in-vessel reflooding and/or reactor cavity flooding has been considered as one of severe accident management strategies and intensive researches to be performed worldwide. This paper provides some results of analytical investigations on the effect of both in- / ex-vessel cooling on the retention of an internally heated molten pool confined in a hemispherical vessel and the related thermal behavior of the vessel wall. For the present analysis, a scale-down reactor vessel for the KSNP reactor design of 1000 MWe (a large dry PWR) is utilized for a reactor vessel. Aluminum oxide melt simulant is also utilized for a real corium pool. An internal power density in the molten pool is determined by a simple scaling analysis that equates the heat flux on the the scale-down vessel wall to that estimated from KSNP. Well-known temperature-dependent boiling heat transfer curves are applied to the in- and ex-vessel cooling boundaries and radiative heat transfer has been only considered in the case of dry in-vessel. MELTPOOL, which is a computational fluid dynamics (CFD) code developed at KAERI, is applied to obtain the time-varying heat flux distribution from a molten pool and the vessel wall temperature distributions with angular positions along the vessel wall. In order to gain further insights on the effectiveness of in- and ex-vessel dual cooling on the in-vessel corium retention, four different boundary conditions has been considered: no water inside the vessel without ex-vessel cooling, water inside the vessel without ex-vessel cooling, no water inside the vessel with ex-vessel cooling, and water inside the vessel with ex-vessel cooling. (authors)

Vacuum vessel for thermonuclear device

International Nuclear Information System (INIS)

Kikuchi, Mitsuru; Kurita, Gen-ichi; Onozuka, Masaki; Suzuki, Masaru.

1997-01-01

Heat of inner walls of a vacuum vessel that receive radiation heat from plasmas by way of first walls is removed by a cooling medium flowing in channels for cooling the inner walls. Nuclear heat generation of constitutional materials of the vacuum vessel caused by fast neutrons and γ rays is removed by a cooling medium flowing in cooling channels disposed in the vacuum vessel. Since the heat from plasmas and the nuclear heat generation are removed separately, the amount of the cooling medium flowing in the channels for cooling inner walls is increased for cooling a great amount of heat from plasmas while the amount of the cooling medium flowing in the channels for cooling the inside of the vacuum vessel is reduced for cooling the small amount of nuclear heat generation. Since the amount of the cooling medium can thus be optimized, the capacity of the facilities for circulating the cooling medium can be reduced. In addition, since the channels for cooling the inner walls and the channels of cooling medium formed in the vacuum vessel are disposed to the inner walls of the vacuum vessel on the side opposite to plasmas, integrity of the channels relative to leakage of the cooling medium can be ensured. (N.H.)

Vacuum vessel for thermonuclear device

Energy Technology Data Exchange (ETDEWEB)

Kikuchi, Mitsuru; Kurita, Gen-ichi [Japan Atomic Energy Research Inst., Tokyo (Japan); Onozuka, Masaki; Suzuki, Masaru

1997-07-31

Heat of inner walls of a vacuum vessel that receive radiation heat from plasmas by way of first walls is removed by a cooling medium flowing in channels for cooling the inner walls. Nuclear heat generation of constitutional materials of the vacuum vessel caused by fast neutrons and {gamma} rays is removed by a cooling medium flowing in cooling channels disposed in the vacuum vessel. Since the heat from plasmas and the nuclear heat generation are removed separately, the amount of the cooling medium flowing in the channels for cooling inner walls is increased for cooling a great amount of heat from plasmas while the amount of the cooling medium flowing in the channels for cooling the inside of the vacuum vessel is reduced for cooling the small amount of nuclear heat generation. Since the amount of the cooling medium can thus be optimized, the capacity of the facilities for circulating the cooling medium can be reduced. In addition, since the channels for cooling the inner walls and the channels of cooling medium formed in the vacuum vessel are disposed to the inner walls of the vacuum vessel on the side opposite to plasmas, integrity of the channels relative to leakage of the cooling medium can be ensured. (N.H.)

Long-wavelength macromolecular crystallography - First successful native SAD experiment close to the sulfur edge

Science.gov (United States)

Aurelius, O.; Duman, R.; El Omari, K.; Mykhaylyk, V.; Wagner, A.

2017-11-01

Phasing of novel macromolecular crystal structures has been challenging since the start of structural biology. Making use of anomalous diffraction of natively present elements, such as sulfur and phosphorus, for phasing has been possible for some systems, but hindered by the necessity to access longer X-ray wavelengths in order to make most use of the anomalous scattering contributions of these elements. Presented here are the results from a first successful experimental phasing study of a macromolecular crystal structure at a wavelength close to the sulfur K edge. This has been made possible by the in-vacuum setup and the long-wavelength optimised experimental setup at the I23 beamline at Diamond Light Source. In these early commissioning experiments only standard data collection and processing procedures have been applied, in particular no dedicated absorption correction has been used. Nevertheless the success of the experiment demonstrates that the capability to extract phase information can be even further improved once data collection protocols and data processing have been optimised.

Tempest in a vessel

International Nuclear Information System (INIS)

Barre, Bertrand

2015-01-01

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

Acute effect of gamma-irradiation on blood vessels of the eye

International Nuclear Information System (INIS)

Matsushima, Hideno

1977-01-01

Radiation injuries to the eye were studied by examining the alteration of fine vasculature of irradiated rabbit eyes, by referring to the usual clinical findings, fundus photography, microangiography, and histological sections. These experiments were performed by using white rabbits whose eyes were locally irradiated with a single dose of 500, 1000, 2000, 3000, and 5000 R of 60 Co-γ-ray. The eyes were observed periodically for a period up to 10 weeks after irradiation, and the following results were obtained: 1) Radiation damage to the blood vessels was maximum around 6 weeks after irradiation for doses less than 2000 R, and the damage appeared to be reversible. 2) When given more than 3000 R irradiation, alteration of the blood vessels continued increasing for 10 weeks with no evidence of recovery. Experimental evidence from these studies supports the hypothesis that the degree of radiation injury of the eye depends highly on the damage to blood vessels serving each structure of the eye. (auth.)

Shuidouchi (Fermented Soybean Fermented in Different Vessels Attenuates HCl/Ethanol-Induced Gastric Mucosal Injury

Directory of Open Access Journals (Sweden)

Huayi Suo

2015-11-01

Full Text Available Shuidouchi (Natto is a fermented soy product showing in vivo gastric injury preventive effects. The treatment effects of Shuidouchi fermented in different vessels on HCl/ethanol-induced gastric mucosal injury mice through their antioxidant effect was determined. Shuidouchi contained isoflavones (daidzein and genistein, and GVFS (glass vessel fermented Shuidouchi had the highest isoflavone levels among Shuidouchi samples fermented in different vessels. After treatment with GVFS, the gastric mucosal injury was reduced as compared to the control mice. The gastric secretion volume (0.47 mL and pH of gastric juice (3.1 of GVFS treated gastric mucosal injury mice were close to those of ranitidine-treated mice and normal mice. Shuidouchi could decrease serum motilin (MTL, gastrin (Gas level and increase somatostatin (SS, vasoactive intestinal peptide (VIP level, and GVFS showed the strongest effects. GVFS showed lower IL-6, IL-12, TNF-α and IFN-γ cytokine levels than other vessel fermented Shuidouchi samples, and these levels were higher than those of ranitidine-treated mice and normal mice. GVFS also had higher superoxide dismutase (SOD, nitric oxide (NO and malonaldehyde (MDA contents in gastric tissues than other Shuidouchi samples. Shuidouchi could raise IκB-α, EGF, EGFR, nNOS, eNOS, Mn-SOD, Gu/Zn-SOD, CAT mRNA expressions and reduce NF-κB, COX-2, iNOS expressions as compared to the control mice. GVFS showed the best treatment effects for gastric mucosal injuries, suggesting that glass vessels could be used for Shuidouchi fermentation in functional food manufacturing.

International Cooperation for the Dismantling of Chooz A Reactor Pressure Vessel

International Nuclear Information System (INIS)

Grenouillet, J.J.; Posivak, E.

2009-01-01

Chooz A is the first PWR that is being decommissioned in France. The main issue that is conditioning the success of the project is the Reactor Pressure Vessel (RPV) and Reactor Vessel Internals (RVI) segmentation. Whereas Chooz A is the first and unique RPV and RVI being dismantled in France, there are many similar experiences available in the world. Thus the project team was eager to cooperate with other teams facing or being faced with the same issue. A cooperation programme was established in two separate ways: - Benefiting from experience feedback from completed RPV and RVI dismantling projects, - Looking for synergy with future RPV dismantling projects for activities such as segmentation tools design, qualification and manufacturing for example. This paper describes the implementation of this programme and how the outcome of the cooperation was used for the implementation of Chooz-A RPV and RVI segmentation project. It shows also the limits of such a cooperation. (authors)

Multi-atlas pancreas segmentation: Atlas selection based on vessel structure.

Science.gov (United States)

Karasawa, Ken'ichi; Oda, Masahiro; Kitasaka, Takayuki; Misawa, Kazunari; Fujiwara, Michitaka; Chu, Chengwen; Zheng, Guoyan; Rueckert, Daniel; Mori, Kensaku

2017-07-01

Automated organ segmentation from medical images is an indispensable component for clinical applications such as computer-aided diagnosis (CAD) and computer-assisted surgery (CAS). We utilize a multi-atlas segmentation scheme, which has recently been used in different approaches in the literature to achieve more accurate and robust segmentation of anatomical structures in computed tomography (CT) volume data. Among abdominal organs, the pancreas has large inter-patient variability in its position, size and shape. Moreover, the CT intensity of the pancreas closely resembles adjacent tissues, rendering its segmentation a challenging task. Due to this, conventional intensity-based atlas selection for pancreas segmentation often fails to select atlases that are similar in pancreas position and shape to those of the unlabeled target volume. In this paper, we propose a new atlas selection strategy based on vessel structure around the pancreatic tissue and demonstrate its application to a multi-atlas pancreas segmentation. Our method utilizes vessel structure around the pancreas to select atlases with high pancreatic resemblance to the unlabeled volume. Also, we investigate two types of applications of the vessel structure information to the atlas selection. Our segmentations were evaluated on 150 abdominal contrast-enhanced CT volumes. The experimental results showed that our approach can segment the pancreas with an average Jaccard index of 66.3% and an average Dice overlap coefficient of 78.5%. Copyright © 2017 Elsevier B.V. All rights reserved.

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)

Analyses of a steel containment vessel with an outer contact structure under severe internal overpressurization conditions

International Nuclear Information System (INIS)

Porter, V.L.

1994-01-01

Many Mark-I and Mark-II BWR plants are designed with a steel vessel as the primary containment. Typically, the steel containment vessel (SCV) is enclosed within a reinforced concrete shield building with only a small gap (74-90 mm) separating the two structures. This paper describes finite element analyses performed to evaluate the effects of contact and friction between a steel containment vessel and an outer contact structure when the containment vessel is subjected to large internal pressures. These computations were motivated by a joint program on containment integrity involving the Nuclear Power Engineering Corporation (NUPEC) of Japan, the US Nuclear Regulatory Commission (NRC), and Sandia National Laboratories for testing model containments. Under severe accident loading conditions, the steel containment vessel in a typical Mark-I or Mark-II plant may deform under internal pressurization such that it contacts the inner surface of a shield building wall. (Thermal expansion from increasing accident temperatures would also close the gap between the SCV and the shield building, but temperature effects are not considered in these analyses.) The amount and location of contact and the pressure at which it occurs all affect how the combined structure behaves. A preliminary finite element model has been developed to analyze a model of a typical steel containment vessel con-ling into contact with an outer structure. Both the steel containment vessel and the outer contact structure were modelled with axisymmetric shell finite elements. Of particular interest are the influence that the contact structure has on deformation and potential failure modes of the containment vessel. Furthermore, the coefficient of friction between the two structures was varied to study its effects on the behavior of the containment vessel and on the uplift loads transmitted to the contact structure. These analyses show that the material properties of an outer contact structure and the amount

Study on the seismic response of reactor vessel of pool type LMFBR including fluid-structure interaction

International Nuclear Information System (INIS)

Tanimoto, K.; Ito, T.; Fujita, K.; Kurihara, C.; Sawada, Y.; Sakurai, A.

1988-01-01

The paper presents the seismic response of reactor vessel of pool type LMFBR with fluid-structure interaction. The reactor vessel has bottom support arrangement, the same core support system as Super-Phenix in France. Due to the bottom support arrangement, the level of core support is lower than that of the side support arrangement. So, in this reactor vessel, the displacement of the core top tends to increase because of the core's rocking. In this study, we investigated the vibration and seismic response characteristics of the reactor vessel. Therefore, the seismic experiments were carried out using one-eighth scale model and the seismic response including FSI and sloshing were investigated. From this study, the effect of liquid on the vibration characteristics and the seismic response characteristics of reactor vessel were clarified and sloshing characteristics were also clarified. It was confirmed that FEM analysis with FSI can reproduce the seismic behavior of the reactor vessel and is applicable to seismic design of the pool type LMFBR with bottom support arrangement. (author). 5 refs, 14 figs, 2 tabs

Seal analysis technology for reactor pressure vessel

International Nuclear Information System (INIS)

Zheng Liangang; Zhang Liping; Yang Yu; Zang Fenggang

2009-01-01

There is the coolant with radiation, high temperature and high pressure in the reactor pressure vessel (RPV). It is closely correlated to RPV sealing capability whether the whole nuclear system work well or not. The aim of this paper is to study the seal analysis method and technology, such as the pre-tensioning of the bolt, elastoplastic contact and coupled technology of thermal and structure. The 3 D elastoplastic seal analysis method really and generally consider the loads and model the contact problem with friction between the contact plates. This method is easier than the specialized seal program and used widely. And it is more really than the 2 D seal analysis method. This 3 D elastoplastic seal analysis method has been successfully used in the design and analysis of RPV. (authors)

Experimental investigations on vessel-hole ablation during severe accidents

International Nuclear Information System (INIS)

Sehgal, B.R.; Dinh, T.N.; Green, J.A.; Paladino, D.

1997-12-01

This report presents experimental results, and subsequent analyses, of scaled reactor pressure vessel (RPV) failure site ablation tests conducted at the Royal Institute of Technology, Division of Nuclear Power Safety (RIT/NPS). The goal of the test program is to reduce the uncertainty level associated with the phase-change-ablation process, and, thus, improve the characterization of the melt discharge loading on the containment. In a series of moderate temperature experiments, the corium melt is simulated by the binary oxide CaO-B 2 O 3 or the binary eutectic and non-eutectic salts NaNO 3 -KNO 3 , while the RPV head steel is represented by a Pb, Sn or metal alloys plate. A complementary set of experiments was conducted at lower temperatures, using water as melt and salted ice as plate material. These experiments scale well to the postulated prototypical conditions. The multidimensional code HAMISA, developed at RIT/NPS, is employed to analyze the experiments with good pre- and post-test predictions. The effects of melt viscosity and crust surface roughness, along with failure site entrance and exit frictional losses on the ablation characteristics are investigated. Theoretical concept was proposed to describe physical mechanisms which govern the vessel-hole ablation process during core melt discharge from RPV. Experimental data obtained from hole ablation tests and separate-effect tests performed at RIT/NPS were used to validate component physical models of the HAMISA code. It is believed that the hole ablation phenomenology is quite well understood. Detailed description of experiments and experimental data, as well as results of analyses are provided in the appendixes

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

International Nuclear Information System (INIS)

Vepa, K.; Sterbentz, W.H.

1981-01-01

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

JSC technician checks STS-44 DSO 316 bioreactor and rotating wall vessel hdwr

Science.gov (United States)

1991-01-01

JSC technician Tacey Prewitt checks the progress on a bioreactor experiment in JSC's Life Sciences Laboratory Bldg 37 biotechnology laboratory. Similar hardware is scheduled for testing aboard Atlantis, Orbiter Vehicle (OV) 104, during STS-44. Detailed Supplementary Objective (DSO) 316 Bioreactor/Flow and Particle Trajectory in Microgravity will checkout the rotating wall vessel hardware and hopefully will confirm researchers' theories and calculations about how flow fields work in space. Plastic beads of various sizes rather than cell cultures are being flown in the vessel for the STS-44 test.

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

Carotid Intraplaque Hemorrhage Imaging with Quantitative Vessel Wall T1 Mapping: Technical Development and Initial Experience.

Science.gov (United States)

Qi, Haikun; Sun, Jie; Qiao, Huiyu; Chen, Shuo; Zhou, Zechen; Pan, Xinlei; Wang, Yishi; Zhao, Xihai; Li, Rui; Yuan, Chun; Chen, Huijun

2018-04-01

Purpose To develop a three-dimensional (3D) high-spatial-resolution time-efficient sequence for use in quantitative vessel wall T1 mapping. Materials and Methods A previously described sequence, simultaneous noncontrast angiography and intraplaque hemorrhage (SNAP) imaging, was extended by introducing 3D golden angle radial k-space sampling (GOAL-SNAP). Sliding window reconstruction was adopted to reconstruct images at different inversion delay times (different T1 contrasts) for voxelwise T1 fitting. Phantom studies were performed to test the accuracy of T1 mapping with GOAL-SNAP against a two-dimensional inversion recovery (IR) spin-echo (SE) sequence. In vivo studies were performed in six healthy volunteers (mean age, 27.8 years ± 3.0 [standard deviation]; age range, 24-32 years; five male) and five patients with atherosclerosis (mean age, 66.4 years ± 5.5; range, 60-73 years; five male) to compare T1 measurements between vessel wall sections (five per artery) with and without intraplaque hemorrhage (IPH). Statistical analyses included Pearson correlation coefficient, Bland-Altman analysis, and Wilcoxon rank-sum test with data permutation by subject. Results Phantom T1 measurements with GOAL-SNAP and IR SE sequences showed excellent correlation (R 2 = 0.99), with a mean bias of -25.8 msec ± 43.6 and a mean percentage error of 4.3% ± 2.5. Minimum T1 was significantly different between sections with IPH and those without it (mean, 371 msec ± 93 vs 944 msec ± 120; P = .01). Estimated T1 of normal vessel wall and muscle were 1195 msec ± 136 and 1117 msec ± 153, respectively. Conclusion High-spatial-resolution (0.8 mm isotropic) time-efficient (5 minutes) vessel wall T1 mapping is achieved by using the GOAL-SNAP sequence. This sequence may yield more quantitative reproducible biomarkers with which to characterize IPH and monitor its progression. © RSNA, 2017.

Plasma discharge in ferritic first wall vacuum vessel of the Hitachi Tokamak HT-2

International Nuclear Information System (INIS)

Abe, Mitsushi; Nakayama, Takeshi; Asano, Katsuhiko; Otsuka, Michio

1997-01-01

A tokamak discharge with ferritic material first wall was tried successfully. The Hitachi Tokamak HT-2 had a stainless steel SUS304 vacuum vessel and modified to have a ferritic plate first wall for experiments to investigate the possibility of ferritic material usage in magnetic fusion devices. The achieved vacuum pressure and times used for discharge cleaning was roughly identical with the stainless steel first wall or the original HT-2. We concluded that ferritic material vacuum vessel is possible for tokamaks. (author)

Vacuum vessel for thermonuclear device

International Nuclear Information System (INIS)

Kikuchi, Mitsuru; Nagashima, Keisuke; Suzuki, Masaru; Onozuka, Masaki.

1997-01-01

A vacuum vessel main body and structural members at the inside and the outside of the vacuum vessel main body are constituted by structural materials activated by irradiation of neutrons from plasmas such as stainless steels. Shielding members comprising tungsten or molybdenum are disposed on the surface of the vacuum vessel main body and the structural members of the inside and the outside of the main body. The shielding members have a function also as first walls or a seat member for the first walls. Armor tiles may be disposed to the shielding members. The shielding members and the armor tiles are secured to a securing seat member disposed, for example, to an inner plate of the vacuum vessel main body by bolts. Since the shielding members are disposed, it is not necessary to constitute the vacuum vessel main body and the structural members at the inside and the outside thereof by using a low activation material which is less activated, such as a titanium alloy. (I.N.)

Vacuum vessel for thermonuclear device

Energy Technology Data Exchange (ETDEWEB)

Kikuchi, Mitsuru; Nagashima, Keisuke [Japan Atomic Energy Research Inst., Tokyo (Japan); Suzuki, Masaru; Onozuka, Masaki

1997-07-11

A vacuum vessel main body and structural members at the inside and the outside of the vacuum vessel main body are constituted by structural materials activated by irradiation of neutrons from plasmas such as stainless steels. Shielding members comprising tungsten or molybdenum are disposed on the surface of the vacuum vessel main body and the structural members of the inside and the outside of the main body. The shielding members have a function also as first walls or a seat member for the first walls. Armor tiles may be disposed to the shielding members. The shielding members and the armor tiles are secured to a securing seat member disposed, for example, to an inner plate of the vacuum vessel main body by bolts. Since the shielding members are disposed, it is not necessary to constitute the vacuum vessel main body and the structural members at the inside and the outside thereof by using a low activation material which is less activated, such as a titanium alloy. (I.N.)