50 CFR 648.4 - Vessel permits.

Science.gov (United States)

2010-10-01

... carrying passengers for hire. (8) Atlantic bluefish vessels. (i) Commercial. Any vessel of the United... lands Atlantic bluefish in or from the EEZ in excess of the recreational possession limit specified at § 648.164 must have been issued and carry on board a valid commercial bluefish vessel permit. (ii) Party...

Review of high thickness welding analysis using SYSWELD for a fusion grade reactor

Energy Technology Data Exchange (ETDEWEB)

Prakash, Ravi, E-mail: prakash@ipr.res.in; Gangradey, Ranjana, E-mail: ranjana@ipr.res.in

2013-10-15

Vacuum vessel and Cryostat for a fusion grade machine are massive structures involving fabrication of chambers with high thickness, about thickness up to 60 mm or more, made of special grade steels. Such machines require accurate planning of welding as the distortions and tolerance levels are stringent. Vacuum vessel of ITER has “D” shaped profile and is toroidal double walled huge steel cage of about 6 m width and 19 m diameter, and the Cryostat of 30 m height and width. The huge vacuum chamber will be fabricated in various parts/sectors due to huge size and then welded with countless weld joints to give the final components. High thickness welding of vacuum vessel is considered to be one of the most important elements in building a reactor of fusion grade due to large ineluctable distortions of welded parts after welding process as it is not easy to correct the large deformations after the welding process and finally the corrections are very expensive. The present paper demonstrates results of welding simulation done using SYSWELD software. Simulation results are of review studies of identified welding process like MIG, MAG, NG-TIG, TIG and EBW for welding large structural D shaped vacuum vessel profile as a case study. Simulation has carried out for SS316LN in clamped as well as unclamped condition for a distortion tolerance of ±2 mm with various weld factors and the local–global approach.

Review of high thickness welding analysis using SYSWELD for a fusion grade reactor

International Nuclear Information System (INIS)

Prakash, Ravi; Gangradey, Ranjana

2013-01-01

Vacuum vessel and Cryostat for a fusion grade machine are massive structures involving fabrication of chambers with high thickness, about thickness up to 60 mm or more, made of special grade steels. Such machines require accurate planning of welding as the distortions and tolerance levels are stringent. Vacuum vessel of ITER has “D” shaped profile and is toroidal double walled huge steel cage of about 6 m width and 19 m diameter, and the Cryostat of 30 m height and width. The huge vacuum chamber will be fabricated in various parts/sectors due to huge size and then welded with countless weld joints to give the final components. High thickness welding of vacuum vessel is considered to be one of the most important elements in building a reactor of fusion grade due to large ineluctable distortions of welded parts after welding process as it is not easy to correct the large deformations after the welding process and finally the corrections are very expensive. The present paper demonstrates results of welding simulation done using SYSWELD software. Simulation results are of review studies of identified welding process like MIG, MAG, NG-TIG, TIG and EBW for welding large structural D shaped vacuum vessel profile as a case study. Simulation has carried out for SS316LN in clamped as well as unclamped condition for a distortion tolerance of ±2 mm with various weld factors and the local–global approach

Creep of A508/533 Pressure Vessel Steel

Energy Technology Data Exchange (ETDEWEB)

Richard Wright

2014-08-01

ABSTRACT Evaluation of potential Reactor Pressure Vessel (RPV) steels has been carried out as part of the pre-conceptual Very High Temperature Reactor (VHTR) design studies. These design studies have generally focused on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Initially, three candidate materials were identified by this process: conventional light water reactor (LWR) RPV steels A508 and A533, 2¼Cr-1Mo in the annealed condition, and Grade 91 steel. The low strength of 2¼Cr-1Mo at elevated temperature has eliminated this steel from serious consideration as the VHTR RPV candidate material. Discussions with the very few vendors that can potentially produce large forgings for nuclear pressure vessels indicate a strong preference for conventional LWR steels. This preference is based in part on extensive experience with forging these steels for nuclear components. It is also based on the inability to cast large ingots of the Grade 91 steel due to segregation during ingot solidification, thus restricting the possible mass of forging components and increasing the amount of welding required for completion of the RPV. Grade 91 steel is also prone to weld cracking and must be post-weld heat treated to ensure adequate high-temperature strength. There are also questions about the ability to produce, and very importantly, verify the through thickness properties of thick sections of Grade 91 material. The availability of large components, ease of fabrication, and nuclear service experience with the A508 and A533 steels strongly favor their use in the RPV for the VHTR. Lowering the gas outlet temperature for the VHTR to 750°C from 950 to 1000°C, proposed in early concept studies, further strengthens the justification for this material selection. This steel is allowed in the ASME Boiler and Pressure Vessel Code for nuclear service up to 371°C (700°F); certain excursions above that temperature are

A specialized vessel for specialized work

Energy Technology Data Exchange (ETDEWEB)

De Vries, J.; De Beenhouwer, S.; Strong, R.; Eversdijk, B. [Tideway BV (Netherlands); Breen, L. [Petro-Canada, St. John' s, NF (Canada)

2003-07-01

This paper describes the challenges associated with the construction and development of the Terra Nova Field, situated on the Grand Banks offshore Newfoundland. In particular, the paper describes how pipelines were protected and stabilized in this harsh environment where the presence of icebergs is common in the springtime. Pipelines were covered with a berm of well-graded quarry stone. The job was executed by the specialized dynamically positioned fall-pipe vessel (DPFPV) called Seahorse. Tideway Marine of the Netherlands operates the Seahorse, which has 2 storage bunkers to carry a total of 18,000 metric tons of rock. Conveyor belts transport the stones from the storage bunkers to a fall-pipe, an open tubular hanging under the vessel which allows the rocks to be dumped in water depths ranging from 20 metres to 1,000 metres. This paper demonstrates that rock dumping is an effective and flexible solution to stabilize and protect pipelines. Dimensions and configuration of the rock can be optimized according to unique specifications of a project. The insulation properties of the rock reduce the amount of insulation foam required for hot flowlines. 2 refs., 4 tabs., 6 figs.

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

33 CFR 155.1045 - Response plan requirements for vessels carrying oil as a secondary cargo.

Science.gov (United States)

2010-07-01

... PREVENTION REGULATIONS FOR VESSELS Tank Vessel Response Plans for Oil § 155.1045 Response plan requirements... actions. (4) The organizational structure that will be used to manage the response actions. This structure... with government agencies; (v) Spill response operations; (vi) Planning; (vii) Logistics support; and...

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

Thick SS316 materials TIG welding development activities towards advanced fusion reactor vacuum vessel applications

Science.gov (United States)

Kumar, B. Ramesh; Gangradey, R.

2012-11-01

Advanced fusion reactors like ITER and up coming Indian DEMO devices are having challenges in terms of their materials design and fabrication procedures. The operation of these devices is having various loads like structural, thermo-mechanical and neutron irradiation effects on major systems like vacuum vessel, divertor, magnets and blanket modules. The concept of double wall vacuum vessel (VV) is proposed in view of protecting of major reactor subsystems like super conducting magnets, diagnostic systems and other critical components from high energy 14 MeV neutrons generated from fusion plasma produced by D-T reactions. The double walled vacuum vessel is used in combination with pressurized water circulation and some special grade borated steel blocks to shield these high energy neutrons effectively. The fabrication of sub components in VV are mainly used with high thickness SS materials in range of 20 mm- 60 mm of various grades based on the required protocols. The structural components of double wall vacuum vessel uses various parts like shields, ribs, shells and diagnostic vacuum ports. These components are to be developed with various welding techniques like TIG welding, Narrow gap TIG welding, Laser welding, Hybrid TIG laser welding, Electron beam welding based on requirement. In the present paper the samples of 20 mm and 40 mm thick SS 316 materials are developed with TIG welding process and their mechanical properties characterization with Tensile, Bend tests and Impact tests are carried out. In addition Vickers hardness tests and microstructural properties of Base metal, Heat Affected Zone (HAZ) and Weld Zone are done. TIG welding application with high thick SS materials in connection with vacuum vessel requirements and involved criticalities towards welding process are highlighted.

Thick SS316 materials TIG welding development activities towards advanced fusion reactor vacuum vessel applications

International Nuclear Information System (INIS)

Kumar, B Ramesh; Gangradey, R

2012-01-01

Advanced fusion reactors like ITER and up coming Indian DEMO devices are having challenges in terms of their materials design and fabrication procedures. The operation of these devices is having various loads like structural, thermo-mechanical and neutron irradiation effects on major systems like vacuum vessel, divertor, magnets and blanket modules. The concept of double wall vacuum vessel (VV) is proposed in view of protecting of major reactor subsystems like super conducting magnets, diagnostic systems and other critical components from high energy 14 MeV neutrons generated from fusion plasma produced by D-T reactions. The double walled vacuum vessel is used in combination with pressurized water circulation and some special grade borated steel blocks to shield these high energy neutrons effectively. The fabrication of sub components in VV are mainly used with high thickness SS materials in range of 20 mm- 60 mm of various grades based on the required protocols. The structural components of double wall vacuum vessel uses various parts like shields, ribs, shells and diagnostic vacuum ports. These components are to be developed with various welding techniques like TIG welding, Narrow gap TIG welding, Laser welding, Hybrid TIG laser welding, Electron beam welding based on requirement. In the present paper the samples of 20 mm and 40 mm thick SS 316 materials are developed with TIG welding process and their mechanical properties characterization with Tensile, Bend tests and Impact tests are carried out. In addition Vickers hardness tests and microstructural properties of Base metal, Heat Affected Zone (HAZ) and Weld Zone are done. TIG welding application with high thick SS materials in connection with vacuum vessel requirements and involved criticalities towards welding process are highlighted.

33 CFR 157.450 - Maneuvering and vessel status information.

Science.gov (United States)

2010-07-01

... SECURITY (CONTINUED) POLLUTION RULES FOR THE PROTECTION OF THE MARINE ENVIRONMENT RELATING TO TANK VESSELS CARRYING OIL IN BULK Interim Measures for Certain Tank Vessels Without Double Hulls Carrying Petroleum Oils...

46 CFR 178.325 - Intact stability requirements for a sailing vessel.

Science.gov (United States)

2010-10-01

... weathertight deck, such as open boats; (4) A vessel that carries more than 49 passengers; (5) A sailing school vessel that carries a combined total of six or more sailing school students or instructors; (6) A vessel... whether the vessel has adequate stability and satisfactory handling characteristics under sail for...

In-service inspection robot for PFBR main vessel- concept

Energy Technology Data Exchange (ETDEWEB)

Rajendran, S; Ramakumar, M S [Bhabha Atomic Research Centre, Mumbai (India). Div. of Remote Handling and Robotics

1994-12-31

In-service inspection (ISI) of critical components in a nuclear reactor is one of the foremost and important tasks which reveals the state of health of the system, thereby ensuring the safety of the plant, personnel and environment. Prototype Fast Breeder Reactor (PFBR) is designed as a pool type reactor. A safety vessel is provided in the design which envelopes the main reactor vessel. The ISI of the main vessel is mandatory and will be carried out by a robot which will operate on this annular gap. The design of the robot is such that it can crawl around the vessel and into the gap at the bottom of the vessel relying on friction grip. The mobile robot will carry a CCTV camera and the inspection technique packages into the interspace, position and orient these to carry out the ISI of the main vessel. The paper discusses about the design features of the robot including the gripping mechanism and the crawling sequence to perform ISI of the reactor vessel. 3 figs.

In-service inspection robot for PFBR main vessel- concept

International Nuclear Information System (INIS)

Rajendran, S.; Ramakumar, M.S.

1994-01-01

In-service inspection (ISI) of critical components in a nuclear reactor is one of the foremost and important tasks which reveals the state of health of the system, thereby ensuring the safety of the plant, personnel and environment. Prototype Fast Breeder Reactor (PFBR) is designed as a pool type reactor. A safety vessel is provided in the design which envelopes the main reactor vessel. The ISI of the main vessel is mandatory and will be carried out by a robot which will operate on this annular gap. The design of the robot is such that it can crawl around the vessel and into the gap at the bottom of the vessel relying on friction grip. The mobile robot will carry a CCTV camera and the inspection technique packages into the interspace, position and orient these to carry out the ISI of the main vessel. The paper discusses about the design features of the robot including the gripping mechanism and the crawling sequence to perform ISI of the reactor vessel. 3 figs

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Evaluation of creep damage due to stress relaxation in SA533 grade B class 1 and SA508 class 3 pressure vessel steels

International Nuclear Information System (INIS)

Hoffmann, C.L.; Urko, W.

1993-01-01

Creep damage can result from stress relaxation of residual stresses in components when exposed to high temperature thermal cycles. Pressure vessels, such as the reactor vessel of the modular high-temperature gas reactor (MHTGR), which normally operate at temperatures well below the creep range can develop relatively high residual stresses in high stress locations. During short term excursions to elevated-temperatures, creep damage can be produced by the loadings on the vessel. In addition, residual stresses will relax out, causing greater creep damage in the pressure vessel material than might otherwise be calculated. The evaluation described in this paper assesses the magnitude of the creep damage due to relaxation of residual stresses resulting from short term exposure of the pressure vessel material to temperatures in the creep range. Creep relaxation curves were generated for SA533 Grade B, Class 1 and SA508 Class 3 pressure vessel steels using finite element analysis of a simple uniaxial truss loaded under constant strain conditions to produce an initial axial stress equal to 1.25 times the material yield strength at temperature. The strain is held constant for 1000 hours at prescribed temperatures from 700 F to 1000 F. The material creep law is used to calculate the relaxed stress for each time increment. The calculated stress relaxation versus time curves are compared with stress relaxation test data. Creep damage fractions are calculated by integrating the stress relaxation versus time curves and performing a linear creep damage summation using the minimum stress to rupture curves at the respective relaxation temperatures. Cumulative creep damage due to stress relaxation as a function of time and temperature is derived from the linear damage summation

33 CFR 157.420 - Vessel specific watch policy and procedures.

Science.gov (United States)

2010-07-01

... SECURITY (CONTINUED) POLLUTION RULES FOR THE PROTECTION OF THE MARINE ENVIRONMENT RELATING TO TANK VESSELS CARRYING OIL IN BULK Interim Measures for Certain Tank Vessels Without Double Hulls Carrying Petroleum Oils...

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

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

International Nuclear Information System (INIS)

Aoki, S.; Ikeda, K.

1978-01-01

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

Life extension of the BR2 aluminium vessel

International Nuclear Information System (INIS)

Koonen, E.; Fabry, A.; Chaouadi, R.; Verwerft, M.; Raedt, C. de; Winckel, S. van; Wacquier, W.; Dadoumont, J.; Verwimp, A.

2000-01-01

The BR2 reactor has recently undergone a major refurbishment comprising the replacement of all vessel internals. The vessel itself however was not replaced. An important requalification programme has been executed to prove that the vessel would remain fit during the contemplated life extension period of BR2. Representative material samples could be obtained from the shroud surrounding the vessel. A comprehensive in-service inspection was carried out and a vessel surveillance programme has been established. (author)

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

Commissioning result of the KSTAR in-vessel cryo-pump

Energy Technology Data Exchange (ETDEWEB)

Chang, Y. B.; Lee, H. J.; Park, Y.M. [National Fusion Research Institute, Daejeon (Korea, Republic of); and others

2013-12-15

KSTAR in-vessel cryo-pump has been installed in the vacuum vessel top and bottom side with up-down symmetry for the better plasma density control in the D-shape H-mode. The cryogenic helium lines of the in-vessel cryo-pump are located at the vertical positions from the vacuum vessel torus center 2,000 mm. The inductive electrical potential has been optimized to reduce risk of electrical breakdown during plasma disruption. In-vessel cryo-pump consists of three parts of coaxial circular shape components; cryo-panel, thermal shield and particle shield. The cryo-panel is cooled down to below 4.5 K. The cryo-panel and thermal shields were made by Inconel 625 tube for higher mechanical strength. The thermal shields and their cooling tubes were annealed in air environment to improve the thermal radiation emissivity on the surface. Surface of cryo-panel was electro-polished to minimize the thermal radiation heat load. The in-vessel cryo-pump was pre-assembled on a test bed in 180 degree segment base. The leak test was carried out after the thermal shock between room temperature to LN2 one before installing them into vacuum vessel. Two segments were welded together in the vacuum vessel and final leak test was performed after the thermal shock. Commissioning of the in-vessel cryo-pump was carried out using a temporary liquid helium supply system.

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.)

Inter-vessels in-service inspection of Super-Phenix

International Nuclear Information System (INIS)

Asty, M.; Saglio, R.; Viard, J.; Lerat, B.

1984-01-01

The vessels design of fast breeder reactor Super-Phenix enables inspection during operating time. A self-moving machine -MIR- has been built up especially for that purpose. It is able to carry out visual and ultrasonorous inspection. MIR structure is that of a tetrahedron, all tops of which are fitted with two wheels, as for traction and direction. The wheels are leaning on booth the two vessels. Thanks to a computer-assisted control system, MIR is able to move along in every part of the inter-vessels space. Studies have been carried on at the French Commissariat a l'Energie Atomique, by two Sections of the advanced technologies Service. After outlining MIR working conditions, its main characteristics are described [fr

Studies on the welding of heavy-section ASTM A542 Cl. 1 steel for large-sized pressure vessels

International Nuclear Information System (INIS)

Shimizu, Shigeki; Aota, Toshiichi; Kasahara, Masayuki

1977-01-01

ASTM A 542, Cl. 1 steel was developed and standardized recently, and is excellent in the high temperature strength and toughness as compared with conventionally used A 387, Grade 22 steel, accordingly the application to large pressure vessels is planned. This steel is a low alloy steel, and in case of large thickness, the possibility of cracking in the welded part is large. Also many times of annealing are required for the prevention of welding cracking, the relieving of residual stress, and the softening of hardened portion, but the possibility of cracking during stress-relieving annealing is large. In this study, Tekken type cracking test was carried out by coated electrode welding, and restricted cracking test was carried out by submerged arc welding of the A 542, Cl. 1 steel and A 387, Grade 22 steel, thus the welding cracking property was investigated, and the optimal welding conditions were selected. Also the test of cracking during the stress-relieving annealing of both steels was carried out, and the method of preventing the cracking was studied. The optimal conditions of stress-relieving annealing were selected, and the mechanism of the cracking was clarified. The mechanical properties of the joints welded and stress-relieved under the selected conditions were confirmed. (Kako, I.)

Seismic proving test of PWR reactor containment vessel

International Nuclear Information System (INIS)

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

1987-01-01

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

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

Diabetic Retinopathy Grading by Digital Curvelet Transform

Directory of Open Access Journals (Sweden)

Shirin Hajeb Mohammad Alipour

2012-01-01

Full Text Available One of the major complications of diabetes is diabetic retinopathy. As manual analysis and diagnosis of large amount of images are time consuming, automatic detection and grading of diabetic retinopathy are desired. In this paper, we use fundus fluorescein angiography and color fundus images simultaneously, extract 6 features employing curvelet transform, and feed them to support vector machine in order to determine diabetic retinopathy severity stages. These features are area of blood vessels, area, regularity of foveal avascular zone, and the number of micro-aneurisms therein, total number of micro-aneurisms, and area of exudates. In order to extract exudates and vessels, we respectively modify curvelet coefficients of color fundus images and angiograms. The end points of extracted vessels in predefined region of interest based on optic disk are connected together to segment foveal avascular zone region. To extract micro-aneurisms from angiogram, first extracted vessels are subtracted from original image, and after removing detected background by morphological operators and enhancing bright small pixels, micro-aneurisms are detected. 70 patients were involved in this study to classify diabetic retinopathy into 3 groups, that is, (1 no diabetic retinopathy, (2 mild/moderate nonproliferative diabetic retinopathy, (3 severe nonproliferative/proliferative diabetic retinopathy, and our simulations show that the proposed system has sensitivity and specificity of 100% for grading.

The JET high temperature in-vessel inspection system

International Nuclear Information System (INIS)

Businaro, T.; Cusack, R.; Calbiati, L.; Raimondi, T.

1989-01-01

The JET In-vessel Inspection System (IVIS) has been enhanced for operation under the following nominal conditions: vacuum vessel at 350 degC; vacuum vessel evacuated (∼10 -9 mbar); radiation dose during D-T phase 10 rads. The target resolution of the pictures is 2 mm at 5 m distance and tests on radiation resistance of the IVIS system are being carried out. Since June 1988, the new system is installed in the JET machine and the first inspections of the intire vessel at 250 degC have been satisfactory done. (author). 3 refs.; 6 figs.; 1 tab

46 CFR 35.25-15 - Carrying of excess steam-TB/ALL.

Science.gov (United States)

2010-10-01

... 46 Shipping 1 2010-10-01 2010-10-01 false Carrying of excess steam-TB/ALL. 35.25-15 Section 35.25... § 35.25-15 Carrying of excess steam—TB/ALL. It shall be the duty of the chief engineer of any tank vessel to see that a steam pressure is not carried in excess of that allowed by the certificate of...

Development of Catamaran Fishing Vessel

Directory of Open Access Journals (Sweden)

A. Jamaluddin

2010-11-01

Full Text Available Multihull due to a couple of advantages has been the topic of extensive research work in naval architecture. In this study, a series of investigation of fishing vessel to save fuel energy was carried out at ITS. Two types of ship models, monohull (round bilge and hard chine and catamaran, a boat with two hulls (symmetrical and asymmetrical were developed. Four models were produced physically and numerically, tested (towing tank and simulated numerically (CFD code. The results of the two approaches indicated that the catamaran mode might have drag (resistance smaller than those of monohull at the same displacement. A layout of catamaran fishing vessel, proposed here, indicates the freedom of setting the deck equipments for fishing vessel.

Predictive factors for short gastric vessels division during laparoscopic total fundoplication

Directory of Open Access Journals (Sweden)

Alexandre Chartuni Pereira Teixeira

Full Text Available OBJECTIVE:to determine clinical variables that can predict the need for division of the short gastric vessels (SGV, based on the gastric fundus tension, assessing postoperative outcomes in patients submitted or not to section of these vessels.METHODS: we analyzed data from 399 consecutive patients undergoing laparoscopic fundoplication for gastroesophageal reflux disease (GERD. The section of the SGV was performed according to the surgeon evaluation, based on the fundus tension. Patients were divided into two groups: not requiring SGV section (group A or requiring SGV section (group B.RESULTS: the section was not necessary in 364 (91% patients (Group A and required in 35 (9% patients (Group B. Group B had proportionally more male patients and higher average height. The endoscopic parameters were worse for Group B, with larger hiatal hernias, greater hernias proportion with more than four centimeters, more intense esophagitis, higher proportion of Barrett's esophagus and long Barrett's esophagus. Male gender and grade IV-V esophagitis were considered independent predictors in the multivariate analysis. Transient dysphagia and GERD symptoms were more common in Group B.CONCLUSION: the division of the short gastric vessels is not required routinely, but male gender and grade IV-V esophagitis are independent predictors of the need for section of these vessels.

46 CFR 25.45-1 - Heating and lighting systems on vessels carrying passengers for hire.

Science.gov (United States)

2010-10-01

... UNINSPECTED VESSELS REQUIREMENTS Cooking, Heating, and Lighting Systems § 25.45-1 Heating and lighting systems...) Alcohol, solid, (2) Alcohol, liquid, combustible, (3) Fuel oil, No. 1, No. 2, or No. 3, (4) Kerosene, (5) Wood or, (6) Coal. (b) Heating and lighting systems using alcohol must meet the following requirements...

46 CFR 117.208 - Survival craft-vessels operating on rivers routes.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Survival craft-vessels operating on rivers routes. 117... LIFESAVING EQUIPMENT AND ARRANGEMENTS Number and Type of Survival Craft § 117.208 Survival craft—vessels... vessel certificated to operate on a rivers route in warm water is not required to carry survival craft...

Method to moor an offshore operating vessel

Energy Technology Data Exchange (ETDEWEB)

Flory, J.F.

1983-01-24

A vessel such as a storage vessel is permanently moored, by means such as a yoke pivoted on the forecastle of the vessel, to a mooring leg, e.g. a riser or anchor chain, which is attached to a base located on the ocean floor. Mounted on the vessel is tension exsisting means, for example, counterweights, springs, winches, or the like, operably connected with the mooring leg for applying tension thereto such as by lifting the yoke. The top of the mooring leg is connected to the end of the yoke through a mooring swivel and a gimbaled mooring table or a universal joint. A fluid swivel may be located above the mooring table or about a load-carrying shaft connected to the mooring leg. 8 drawings.

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

International Nuclear Information System (INIS)

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

1986-01-01

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

State of opening the cover and carrying out the checkup of the reactor vessel of the nuclear-powered ship 'Mutsu' by Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

1989-01-01

In the checkup by opening the cover of the reactor vessel of the nuclear-powered ship 'Mutsu', Japan Atomic Energy Research Institute carried out the checkup and maintenance for the reactor proper, control system and primary coolant facilities including the secondary side of steam generators and the pressure balancing valve of the containment vessel. The works were classified into the opening of the reactor, checkup, maintenance and restoration. The opening was begun on August 4, 1988, and finished on December 5. The checkup and maintenance were begun on September 22, and are still continued now. The maximum radiation dose rate on the surfaces of fuel assemblies and control rods and at the positions 1 m distant from them was measured. The results of the checkup of various components are reported. In 290 absorbent rods of control rods, spot corrosion and discoloration were observed, of which the spot corrosion penetrated the walls of 4 rods. Also in 12 fuel rods, spot corrosion was observed near the welded end plugs, but leak was not observed. (K.I.)

Analysis of aging mechanism and management for HTR-PM reactor pressure vessel

International Nuclear Information System (INIS)

Sun Yunxue; Shao Jin

2015-01-01

Reactor pressure vessel is an important part of the reactor pressure boundary, its important degree ranks high in ageing management and life assessment of nuclear power plant. Carrying out systematic aging management to ensure reactor pressure vessel keeping enough safety margins and executing design functions is one of the key factors to guarantee security and stability operation for nuclear power plant during the whole lifetime and prolong life. This paper briefly introduces the structure and aging mechanism of reactor pressure vessel in pressurized water reactor nuclear power plant, and introduces the design principle and structure characteristics of HTR-PM. At the same time, this paper carries out preliminary analysis and exploration. and discusses aging management of HTR-PM reactor pressure vessel. Finally, the advice of carring out aging management for HTR-PM reactor pressure vessel is proposed. (authors)

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

Diffuse high-grade gliomas with H3 K27M mutations carry a dismal prognosis independent of tumor location.

Science.gov (United States)

Karremann, Michael; Gielen, Gerrit H; Hoffmann, Marion; Wiese, Maria; Colditz, Niclas; Warmuth-Metz, Monika; Bison, Brigitte; Claviez, Alexander; van Vuurden, Dannis G; von Bueren, André O; Gessi, Marco; Kühnle, Ingrid; Hans, Volkmar H; Benesch, Martin; Sturm, Dominik; Kortmann, Rolf-Dieter; Waha, Andreas; Pietsch, Torsten; Kramm, Christof M

2018-01-10

The novel entity of "diffuse midline glioma, H3 K27M-mutant" has been defined in the 2016 revision of the World Health Organization (WHO) classification of tumors of the central nervous system (CNS). Tumors of this entity arise in CNS midline structures of predominantly pediatric patients and are associated with an overall dismal prognosis. They are defined by K27M mutations in H3F3A or HIST1H3B/C, encoding for histone 3 variants H3.3 and H3.1, respectively, which are considered hallmark events driving gliomagenesis. Here, we characterized 85 centrally reviewed diffuse gliomas on midline locations enrolled in the nationwide pediatric German HIT-HGG registry regarding tumor site, histone 3 mutational status, WHO grade, age, sex, and extent of tumor resection. We found 56 H3.3 K27M-mutant tumors (66%), 6 H3.1 K27M-mutant tumors (7%), and 23 H3-wildtype tumors (27%). H3 K27M-mutant gliomas shared an aggressive clinical course independent of their anatomic location. Multivariate regression analysis confirmed the significant impact of the H3 K27M mutation as the only independent parameter predictive of overall survival (P = 0.009). In H3 K27M-mutant tumors, neither anatomic midline location nor histopathological grading nor extent of tumor resection had an influence on survival. These results substantiate the clinical significance of considering diffuse midline glioma, H3 K27M-mutant, as a distinct entity corresponding to WHO grade IV, carrying a universally fatal prognosis. © The Author(s) 2017. Published by Oxford University Press on behalf of the Society for Neuro-Oncology. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com

Oscillation of Angiogenesis with Vascular Dropout in Diabetic Retinopathy by VESsel GENeration Analysis (VESGEN)

Science.gov (United States)

Parsons-Wingerter, Patricia; Radbakrishnan, Krisbnan; Vickerman, Mary B.; Kaiser, Peter K.

2010-01-01

PURPOSE. Vascular dropout and angiogenesis are hallmarks of the progression of diabetic retinopathy (DR). However, current evaluation of DR relies on grading of secondary vascular effects, such as microaneurysms and hemorrhages, by clinical examination instead of by evaluation of actual vascular changes. The purpose of this study was to map and quantify vascular changes during progression of DR by VESsel GENeration Analysis (VESGEN). METHODS. In this prospective cross-sectional study, 15 eyes with DR were evaluated with fluorescein angiography (FA) and color fundus photography, and were graded using modified Early Treatment Diabetic Retinopathy Study criteria. FA images were separated by semiautomatic image processing into arterial and venous trees. Vessel length density (L(sub v)), number density (N(sub v)), and diameter (D(sub v)) were analyzed in a masked fashion with VESGEN software. Each vascular tree was automatically segmented into branching generations (G(sub 1)...G(sub 8) or G(sub 9)) by vessel diameter and branching. Vascular remodeling status (VRS) for N(sub v) and L(sub v) was graded 1 to 4 for increasing severity of vascular change. RESULTS. By N(sub v) and L(sub v), VRS correlated significantly with the independent clinical diagnosis of mild to proliferative DR (13/15 eyes). N(sub v) and L(sub v) of smaller vessels (G(sub >=6) increased from VRS1 to VRS2 by 2.4 X and 1.6 X, decreased from VRS2 to VRS3 by 0.4 X and 0.6X, and increased from VRS3 to VRS4 by 1.7 X and 1.5 X (P < 0.01). Throughout DR progression, the density of larger vessels (G(sub 1-5)) remained essentially unchanged, and D(sub v1-5) increased slightly. CONCLUSIONS. Vessel density oscillated with the progression of DR. Alternating phases of angiogenesis/neovascularization and vascular dropout were dominated first by remodeling of arteries and subsequently by veins.

Dynamic stroma reorganization drives blood vessel dysmorphia during glioma growth.

Science.gov (United States)

Mathivet, Thomas; Bouleti, Claire; Van Woensel, Matthias; Stanchi, Fabio; Verschuere, Tina; Phng, Li-Kun; Dejaegher, Joost; Balcer, Marly; Matsumoto, Ken; Georgieva, Petya B; Belmans, Jochen; Sciot, Raf; Stockmann, Christian; Mazzone, Massimiliano; De Vleeschouwer, Steven; Gerhardt, Holger

2017-12-01

Glioma growth and progression are characterized by abundant development of blood vessels that are highly aberrant and poorly functional, with detrimental consequences for drug delivery efficacy. The mechanisms driving this vessel dysmorphia during tumor progression are poorly understood. Using longitudinal intravital imaging in a mouse glioma model, we identify that dynamic sprouting and functional morphogenesis of a highly branched vessel network characterize the initial tumor growth, dramatically changing to vessel expansion, leakage, and loss of branching complexity in the later stages. This vascular phenotype transition was accompanied by recruitment of predominantly pro-inflammatory M1-like macrophages in the early stages, followed by in situ repolarization to M2-like macrophages, which produced VEGF-A and relocate to perivascular areas. A similar enrichment and perivascular accumulation of M2 versus M1 macrophages correlated with vessel dilation and malignancy in human glioma samples of different WHO malignancy grade. Targeting macrophages using anti-CSF1 treatment restored normal blood vessel patterning and function. Combination treatment with chemotherapy showed survival benefit, suggesting that targeting macrophages as the key driver of blood vessel dysmorphia in glioma progression presents opportunities to improve efficacy of chemotherapeutic agents. We propose that vessel dysfunction is not simply a general feature of tumor vessel formation, but rather an emergent property resulting from a dynamic and functional reorganization of the tumor stroma and its angiogenic influences. © 2017 The Authors. Published under the terms of the CC BY 4.0 license.

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

International Nuclear Information System (INIS)

Dorval, C.

1966-05-01

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

Preliminary structural evaluations of the STAR-LM reactor vessel and the support design

International Nuclear Information System (INIS)

Koo, Gyeong-Hoi; Sienicki, James J.; Moisseytsev, Anton

2007-01-01

In this paper, preliminary structural evaluations of the reactor vessel and support design of the STAR-LM (The Secure, Transportable, Autonomous Reactor - Liquid Metal variant), which is a lead-cooled reactor, are carried out with respect to an elevated temperature design and seismic design. For an elevated temperature design, the structural integrity of a direct coolant contact to the reactor vessel is investigated by using a detail structural analysis and the ASME-NH code rules. From the results of the structural analyses and the integrity evaluations, it was found that the design concept of a direct coolant contact to the reactor vessel cannot satisfy the ASME-NH rules for a given design condition. Therefore, a design modification with regards to the thermal barrier is introduced in the STAR-LM design. For a seismic design, detailed seismic time history response analyses for a reactor vessel with a consideration of a fluid-structure interaction are carried out for both a top support type and a bottom support type. And from the results of the hydrodynamic pressure responses, an investigation of the minimum thickness design of the reactor vessel is tentatively carried out by using the ASME design rules

Remote controlled ultrasonic pre-service and in-service inspections of reactor pressure vessels

International Nuclear Information System (INIS)

Mueller, G.

1990-01-01

The first mechanised in-service inspection of the reactor pressure vessel on unit one of Eskom's Koeberg nuclear power station has been carried out. Since 1968 a whole range of manipulators to carry out remote controlled ultrasonic inspections of nuclear power station equipment has been developed. The inspection of a reactor pressure vessel using a central mast manipulator is described. 3 figs., 1 ill

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).

Pressure vessel for a BWR type reactor

International Nuclear Information System (INIS)

Shimamoto, Yoshiharu.

1980-01-01

Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

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

Crack propagation on spherical pressure vessels

International Nuclear Information System (INIS)

Lebey, J.; Roche, R.

1975-01-01

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

Shear strength of end slabs of prestressed concrete nuclear reactor vessels

International Nuclear Information System (INIS)

Reins, J.D.; Quiros, J.L. Jr.; Schnobrich, W.C.; Sozen, M.A.

1976-07-01

The report summarizes the experimental and part of the analytical work carried out in connection with an investigation of the structural strength of prestressed concrete reactor vessels. The project is part of the Prestressed Concrete Reactor Vessel Program of the Oak Ridge National Laboratory sponsored by ERDA. The objective of the current phase of the work is to develop procedures to determine the shear strength of flat end slabs of reactor vessels with penetrations

Application of Response Surface Methodology for Modeling of Postweld Heat Treatment Process in a Pressure Vessel Steel ASTM A516 Grade 70.

Science.gov (United States)

Peasura, Prachya

2015-01-01

This research studied the application of the response surface methodology (RSM) and central composite design (CCD) experiment in mathematical model and optimizes postweld heat treatment (PWHT). The material of study is a pressure vessel steel ASTM A516 grade 70 that is used for gas metal arc welding. PWHT parameters examined in this study included PWHT temperatures and time. The resulting materials were examined using CCD experiment and the RSM to determine the resulting material tensile strength test, observed with optical microscopy and scanning electron microscopy. The experimental results show that using a full quadratic model with the proposed mathematical model is YTS = -285.521 + 15.706X1 + 2.514X2 - 0.004X1(2) - 0.001X2(2) - 0.029X1X2. Tensile strength parameters of PWHT were optimized PWHT time of 5.00 hr and PWHT temperature of 645.75°C. The results show that the PWHT time is the dominant mechanism used to modify the tensile strength compared to the PWHT temperatures. This phenomenon could be explained by the fact that pearlite can contribute to higher tensile strength. Pearlite has an intensity, which results in increased material tensile strength. The research described here can be used as material data on PWHT parameters for an ASTM A516 grade 70 weld.

Vessel architecture in human knee cartilage in children: an in vivo susceptibility-weighted imaging study at 7 T.

Science.gov (United States)

Kolb, Alexander; Robinson, Simon; Stelzeneder, David; Schreiner, Markus; Chiari, Catharina; Windhager, Reinhard; Trattnig, Siegfried; Bohndorf, Klaus

2018-02-26

To evaluate the clinical feasibility of ultrahigh field 7-T SWI to visualize vessels and assess their density in the immature epiphyseal cartilage of human knee joints. 7-T SWI of 12 knees (six healthy volunteers, six patients with osteochondral abnormalities; mean age 10.7 years; 3 female, 9 male) were analysed by two readers, classifying intracartilaginous vessel densities (IVD) in three grades (no vessels, low IVD and high IVD) in defined femoral, tibial and patellar zones. Differences between patients and volunteers, IVDs in different anatomic locations, differences between cartilage overlying osteochondral abnormalities and corresponding normal zones, and differences in age groups were analysed. Interrater reliability showed moderate agreement between the two readers (κ = 0.58, p < 0.001). The comparison of IVDs between patients and volunteers revealed no significant difference (p = 0.706). The difference between zones in the cartilage overlying osteochondral abnormalities to corresponding normal zones showed no significant difference (p = 0.564). IVDs were related to anatomic location, with decreased IVDs in loading areas (p = 0.003). IVD was age dependent, with more vessels present in the younger participants (p = 0.001). The use of SWI in conjunction with ultrahigh field MRI makes the in vivo visualization of vessels in the growing cartilage of humans feasible, providing insights into the role of the vessel network in acquired disturbances. • SWI facilitates in vivo visualization of vessels in the growing human cartilage. • Interrater reliability of the intracartilaginous vessel grading was moderate. • Intracartilaginous vessel densities are dependent on anatomical location and age.

Application of high strength steel to nuclear reactor containment vessel

International Nuclear Information System (INIS)

Susukida, H.; Sato, M.; Takano, G.; Uebayashi, T.; Yoshida, K.

1976-01-01

Nuclear reactor containment vessels are becoming larger in size with the increase in the power generating capacity of nuclear power plants. For example, a containment vessel for a PWR power plant with an output of 1,000 MWe becomes an extremely large one if it is made of the conventional JIS SGV 49 (ASTM A 516 Gr. 70) steel plates less than 38 mm in thickness. In order to design the steel containment vessel within the conventional dimensional range, therefore, it is necessary to use a high strength steel having a higher tensile strength than SGV 49 steel, good weldability and a higher fracture toughness and moreover, possessing satisfactory properties without undergoing post-weld heat treatment. The authors conducted a series of verification tests on high strength steel developed by modifying the ASTM A 543 Grade B Class 1 steel with a view to adopting it as a material for the nuclear reactor containment vessels. As the result of evaluation of the test results from various angles, we confirmed that the high strength steel is quite suitable for the manufacture of nuclear reactor containment vessels. (auth.)

Validated automated ultrasonic inspections of the Sizewell 'B' reactor pressure vessel

International Nuclear Information System (INIS)

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

1992-01-01

Automated ultrasonic inspection was applied extensively during manufacture of the RPV for Sizewell 'B'. This was an important element of the safety case presented at the Sizewell 'B' public enquiry. This requirement reflected concern in the United Kingdom as to the effectiveness and reliability of ultrasonic inspections. By applying automated inspections in addition to the manual ultrasonic inspection carried out by the vessel manufacturer, the overall reliability of the inspection of the vessel would be considerably enhanced. The automated inspections carried out in the manufacturer's workshops were termed 'automated shop inspections' (ASIs). The ASIs were carried out in two contracts: the first to inspect the component forgings of the RPV, the second to inspect the pressure retaining welds. (author)

Oscillation of Angiogenesis with Vascular Dropout in Diabetic Retinopathy by VESsel GENeration Analysis (VESGEN)

Science.gov (United States)

Parsons-Wingerter, Patricia; Radbakrishnan, Krisbnan; Vickerman, Mary B.; Kaiser, Peter K.

2010-01-01

PURPOSE. Vascular dropout and angiogenesis are hallmarks of the progression of diabetic retinopathy (DR). However, current evaluation of DR relies on grading of secondary vascular effects, such as microaneurysms and hemorrhages, by clinical examination instead of by evaluation of actual vascular changes. The purpose of this study was to map and quantify vascular changes during progression of DR by VESsel GENeration Analysis (VESGEN). METHODS. In this prospective cross-sectional study, 15 eyes with DR were evaluated with fluorescein angiography (FA) and color fundus photography, and were graded using modified Early Treatment Diabetic Retinopathy Study criteria. FA images were separated by semiautomatic image processing into arterial and venous trees. Vessel length density (L(sub v)), number density (N(sub v)), and diameter (D(sub v)) were analyzed in a masked fashion with VESGEN software. Each vascular tree was automatically segmented into branching generations (G(sub 1)...G(sub 8) or G(sub 9)) by vessel diameter and branching. Vascular remodeling status (VRS) for N(sub v) and L(sub v) was graded 1 to 4 for increasing severity of vascular change. RESULTS. By N(sub v) and L(sub v), VRS correlated significantly with the independent clinical diagnosis of mild to proliferative DR (13/15 eyes). N(sub v) and L(sub v) of smaller vessels (G(sub >=6) increased from VRS1 to VRS2 by 2.4 X and 1.6 X, decreased from VRS2 to VRS3 by 0.4 X and 0.6X, and increased from VRS3 to VRS4 by 1.7 X and 1.5 X (P dropout were dominated first by remodeling of arteries and subsequently by veins.

Automated grading system for evaluation of ocular redness associated with dry eye.

Science.gov (United States)

Rodriguez, John D; Johnston, Patrick R; Ousler, George W; Smith, Lisa M; Abelson, Mark B

2013-01-01

We have observed that dry eye redness is characterized by a prominence of fine horizontal conjunctival vessels in the exposed ocular surface of the interpalpebral fissure, and have incorporated this feature into the grading of redness in clinical studies of dry eye. To develop an automated method of grading dry eye-associated ocular redness in order to expand on the clinical grading system currently used. Ninety nine images from 26 dry eye subjects were evaluated by five graders using a 0-4 (in 0.5 increments) dry eye redness (Ora Calibra™ Dry Eye Redness Scale [OCDER]) scale. For the automated method, the Opencv computer vision library was used to develop software for calculating redness and horizontal conjunctival vessels (noted as "horizontality"). From original photograph, the region of interest (ROI) was selected manually using the open source ImageJ software. Total average redness intensity (Com-Red) was calculated as a single channel 8-bit image as R - 0.83G - 0.17B, where R, G and B were the respective intensities of the red, green and blue channels. The location of vessels was detected by normalizing the blue channel and selecting pixels with an intensity of less than 97% of the mean. The horizontal component (Com-Hor) was calculated by the first order Sobel derivative in the vertical direction and the score was calculated as the average blue channel image intensity of this vertical derivative. Pearson correlation coefficients, accuracy and concordance correlation coefficients (CCC) were calculated after regression and standardized regression of the dataset. The agreement (both Pearson's and CCC) among investigators using the OCDER scale was 0.67, while the agreement of investigator to computer was 0.76. A multiple regression using both redness and horizontality improved the agreement CCC from 0.66 and 0.69 to 0.76, demonstrating the contribution of vessel geometry to the overall grade. Computer analysis of a given image has 100% repeatability and zero

System for cooling the upper wall of a nuclear reactor vessel

International Nuclear Information System (INIS)

Pailla, Henri; Schaller, Karl; Vidard, Michel.

1974-01-01

A system for cooling the upper wall of the main vessel of a fast neutron reactor is described. This vessel is suspended from an upper shield by the upper wall. It includes coils carrying a coolant which are immersed in an intermediate liquid bathing the wall and contained in a tank integral with the vessel. At least one of the two cooling and intermediate liquids is a liquid metal. The main vessel is contained in a safety vessel, the space between the main and safety vessels is occluded in its upper part by an insulating shield placed under the tank. There is a liquid metal seal between the upper wall and the upper shield under the tank. This system has been specially designed for sodium cooled fast neutron reactors [fr

Research and development of spent fuel shipping casks and the criteria for seagoing vessel carrying casks

International Nuclear Information System (INIS)

Aoki, S.; Ando, Y.

1977-01-01

Considering that the transportation of spent fuel will increase rapidly and extensively in the near future, Japanese Atomic Energy Committee enacted ''Technical Standard for Transportation of Radioactive Materials'' based on ''IAEA Regulation for the Safe Transport of Radioactive Materials 1973 Revised Edition''. Coping with the recommendation of AEC, Atomic Energy Bureau in Science and Technology Agency and other authorities concerned started to review the former ordinances for transportation of radioactive materials and to consolidate a unified system of relevant laws and standards. On the other hand, Atomic Energy Bureau has invested in research and development since ten years ago in order to obtain the data for design and licensing work of spent fuel shipping casks. In those studies some different scale models of a prototype of 80 t in weight have been used to make clear the scale effect at the drop, pucture and fire tests, which are one of the features of Japanese research and development. And also the immersion test in high pressure water up to about 500 bars is now carried out to investigate the integrity of cask body and sealing structure to prevent leakage of radioactive contents to the ambient when the cask falls into deep sea. In Japan, depending on the site conditions of nuclear plants, almost all transportations of unirradiated and spent fuels are done on the sea. Therefore, in order to secure the safety of transportation, the design criteria of the seagoing vessels for exclusive transportation of spent fuel shipping casks, namely full load shipping, has been enacted, which aims to make minimum the probability of sinking at collison, stranding and other unforeseen accidents at sea and also to restrain radiation exposure of the crew as low as possible

46 CFR 180.207 - Survival craft-vessels operating on lakes, bays, and sounds routes.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Survival craft-vessels operating on lakes, bays, and... Survival Craft § 180.207 Survival craft—vessels operating on lakes, bays, and sounds routes. (a) Except as... warm water is not required to carry survival craft. (d) A vessel certificated to operate on lakes, bays...

Seismic transient analysis of a containment vessel with penetrations

International Nuclear Information System (INIS)

Dahlke, H.J.; Weiner, E.O.

1979-12-01

A linear transient analysis of the FFTF containment vessel was conducted with STAGS to justify the load levels used for the seismic qualification testing of the heating and ventiliation valve operators. The modeling consists of a thin axisymmetric shell for the containment vessel with four penetrations characterized by linear and rotational inertias as well as attachment characteristics to the shell. Motions considered are horizontal, rocking and vertical input to the base, and the solution is carried out by direct integration. Results show that the test levels and the approximate analyses considered are conservative. Response spectra for some containment vessel penetrations applicable to the model are presented

Biomechanical, Physiological, and Agility Performance of Soldiers Carrying Loads: A Comparison of the Modular Lightweight Load Carrying Equipment and a Lightning Packs, LLC, Prototype

Science.gov (United States)

2016-12-27

angle, hip angle, and sagittal plane hip moments. In terms of energy harvesting and production during walking, the current weight penalty of carrying...MODULAR LIGHTWEIGHT LOAD CARRYING EQUIPMENT) HUMAN FACTORS ENGINEERING U.S. Army Natick Soldier Research, Development and Engineering Center ATTN...pack type and walking speed at a 0% grade. .......................................................35  vii Table 20: Means (SE) of the mean and

Possible research program on a large scale nuclear pressure vessel

International Nuclear Information System (INIS)

1983-01-01

The nuclear pressure vessel structural integrity is actually one of the main items in the nuclear plants safety field. An international study group aimed at investigating the feasibility of a ''possible research program'' on a scale 1:1 LWR pressure vessel. This report presents the study group's work. The different research programs carried out or being carried out in various countries of the European Community are presented (phase I of the study). The main characteristics of the vessel considered for the program and an evaluation of activities required for making them available are listed. Research topic priorities from the different interested countries are summarized in tables (phase 2); a critical review by the study group of the topic is presented. Then, proposals for possible experimental programs and combination of these programs are presented, only as examples of possible useful research activities. The documents pertaining to the results of phase I inquiry performed by the study group are reported in the appendix

In-service ultrasonic inspection of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Prepechal, J.; Sulc, J.

1982-01-01

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

A Nordic approach to impact assessment of accidents with nuclear-propelled vessels

International Nuclear Information System (INIS)

Reistad, O.; Hustveit, S.; Palsson, S.E.; Hoe, S.; Lahtinen, J.

2012-11-01

The MareNuc project has identified the parameters in a graded approach to impact assessment for marine nuclear reactors. The graded approach is founded on the following elements: 1) More detailed understanding of previous accidents in nuclear-propelled vessels (initiating events, accident developments, release fractions), including release mechanisms (radionuclide retention in vessel construction); 2) Bench-marking of release scenarios using modelling tools applied in the Nordic countries, in addition to demonstration of generally accessible and free software developed by the IAEA; 3) Other systematic approaches to safety assessments of vessel port calls, and to the design and maintenance of emergency preparedness systems; More specifically, increased emphasis compared to earlier analysis after the Kursk accident is given to the engineered vessel barriers. Relevant standards from impact assessments for commercial nuclear power plants have been identified, such as from the NUREG series. The Nordic approaches to safety evaluation, impact assessments and emergency preparedness organisation was also reported as part of the project. The Canadian approach for international port calls was carefully reported and assessed as part of the project, and commended for its broad and comprehensive approach to reactor and vessel design for the nationalities involved, to the design and maintenance of emergency preparedness systems, and the well-structured and broad cooperation between civilian and military institutions. This approach goes beyond the current approach in the Nordic countries, also in the case of Norway, which experience regular port calls from allied nuclear navies. The overall result is a broader understanding in the Nordic countries for the importance of the various parameters for impact assessment of releases from marine reactors, and to the design and maintenance of an emergency preparedness organisation without detailed knowledge of the installation in question

A Nordic approach to impact assessment of accidents with nuclear-propelled vessels

Energy Technology Data Exchange (ETDEWEB)

Reistad, O. [Institute for Energy Technology, Kjeller (Norway); Hustveit, S. [Norwegian Radiation Protection Authority, Oesteraes (Norway); Palsson, S.E. [Icelandic Radiation Safety Authority, Reykjavik (Iceland); Hoe, S. [Danish Emergency Management Agency, Birkeroed (Denmark); Lahtinen, J. [STUK, Helsinki (Finland)

2012-11-15

The MareNuc project has identified the parameters in a graded approach to impact assessment for marine nuclear reactors. The graded approach is founded on the following elements: 1) More detailed understanding of previous accidents in nuclear-propelled vessels (initiating events, accident developments, release fractions), including release mechanisms (radionuclide retention in vessel construction); 2) Bench-marking of release scenarios using modelling tools applied in the Nordic countries, in addition to demonstration of generally accessible and free software developed by the IAEA; 3) Other systematic approaches to safety assessments of vessel port calls, and to the design and maintenance of emergency preparedness systems; More specifically, increased emphasis compared to earlier analysis after the Kursk accident is given to the engineered vessel barriers. Relevant standards from impact assessments for commercial nuclear power plants have been identified, such as from the NUREG series. The Nordic approaches to safety evaluation, impact assessments and emergency preparedness organisation was also reported as part of the project. The Canadian approach for international port calls was carefully reported and assessed as part of the project, and commended for its broad and comprehensive approach to reactor and vessel design for the nationalities involved, to the design and maintenance of emergency preparedness systems, and the well-structured and broad cooperation between civilian and military institutions. This approach goes beyond the current approach in the Nordic countries, also in the case of Norway, which experience regular port calls from allied nuclear navies. The overall result is a broader understanding in the Nordic countries for the importance of the various parameters for impact assessment of releases from marine reactors, and to the design and maintenance of an emergency preparedness organisation without detailed knowledge of the installation in question

Carotid artery stenosis: Performance of advanced vessel analysis software in evaluating CTA

International Nuclear Information System (INIS)

Tsiflikas, Ilias; Biermann, Christina; Thomas, Christoph; Ketelsen, Dominik; Claussen, Claus D.; Heuschmid, Martin

2012-01-01

Objectives: The aim of this study was to evaluate time efficiency and diagnostic reproducibility of an advanced vessel analysis software for diagnosis of carotid artery stenosis. Material and methods: 40 patients with suspected carotid artery stenosis received head and neck DE-CTA as part of their pre-interventional workup. Acquired data were evaluated by 2 independent radiologists. Stenosis grading was performed by MPR eyeballing with freely adjustable MPRs and with a preliminary prototype of the meanwhile available client-server and advanced visualization software syngo.via CT Vascular (Siemens Healthcare, Erlangen, Germany). Stenoses were graded according to the following 5 categories: I: 0%, II: 1–50%, III: 51–69%, IV: 70–99% and V: total occlusion. Furthermore, time to diagnosis for each carotid artery was recorded. Results: Both readers achieved very good specificity values and good respectively very good sensitivity values without significant differences between both reading methods. Furthermore, there was a very good correlation between both readers for both reading methods without significant differences (kappa value: standard image interpretation k = 0.809; advanced vessel analysis software k = 0.863). Using advanced vessel analysis software resulted in a significant time saving (p < 0.0001) for both readers. Time to diagnosis could be decreased by approximately 55%. Conclusions: Advanced vessel analysis application CT Vascular of the new imaging software syngo.via (Siemens Healthcare, Forchheim, Germany) provides a high rate of reproducibility in assessment of carotid artery stenosis. Furthermore a significant time saving in comparison to standard image interpretation is achievable

Carotid artery stenosis: Performance of advanced vessel analysis software in evaluating CTA

Energy Technology Data Exchange (ETDEWEB)

Tsiflikas, Ilias, E-mail: ilias.tsiflikas@med.uni-tuebingen.de [University Hospital of Tuebingen, Diagnostic and Interventional Radiology, Hoppe-Seyler-Str. 3, 72076 Tuebingen (Germany); Biermann, Christina, E-mail: christina.biermann@siemens.com [University Hospital of Tuebingen, Diagnostic and Interventional Radiology, Hoppe-Seyler-Str. 3, 72076 Tuebingen (Germany); Siemens AG, Siemens Healthcare Consulting, Allee am Röthelheimpark 3A, 91052 Erlangen (Germany); Thomas, Christoph, E-mail: christoph.thomas@med.uni-tuebingen.de [University Hospital of Tuebingen, Diagnostic and Interventional Radiology, Hoppe-Seyler-Str. 3, 72076 Tuebingen (Germany); Ketelsen, Dominik, E-mail: dominik.ketelsen@med.uni-tuebingen.de [University Hospital of Tuebingen, Diagnostic and Interventional Radiology, Hoppe-Seyler-Str. 3, 72076 Tuebingen (Germany); Claussen, Claus D., E-mail: claus.claussen@med.uni-tuebingen.de [University Hospital of Tuebingen, Diagnostic and Interventional Radiology, Hoppe-Seyler-Str. 3, 72076 Tuebingen (Germany); Heuschmid, Martin, E-mail: martin.heuschmid@med.uni-tuebingen.de [University Hospital of Tuebingen, Diagnostic and Interventional Radiology, Hoppe-Seyler-Str. 3, 72076 Tuebingen (Germany)

2012-09-15

Objectives: The aim of this study was to evaluate time efficiency and diagnostic reproducibility of an advanced vessel analysis software for diagnosis of carotid artery stenosis. Material and methods: 40 patients with suspected carotid artery stenosis received head and neck DE-CTA as part of their pre-interventional workup. Acquired data were evaluated by 2 independent radiologists. Stenosis grading was performed by MPR eyeballing with freely adjustable MPRs and with a preliminary prototype of the meanwhile available client-server and advanced visualization software syngo.via CT Vascular (Siemens Healthcare, Erlangen, Germany). Stenoses were graded according to the following 5 categories: I: 0%, II: 1–50%, III: 51–69%, IV: 70–99% and V: total occlusion. Furthermore, time to diagnosis for each carotid artery was recorded. Results: Both readers achieved very good specificity values and good respectively very good sensitivity values without significant differences between both reading methods. Furthermore, there was a very good correlation between both readers for both reading methods without significant differences (kappa value: standard image interpretation k = 0.809; advanced vessel analysis software k = 0.863). Using advanced vessel analysis software resulted in a significant time saving (p < 0.0001) for both readers. Time to diagnosis could be decreased by approximately 55%. Conclusions: Advanced vessel analysis application CT Vascular of the new imaging software syngo.via (Siemens Healthcare, Forchheim, Germany) provides a high rate of reproducibility in assessment of carotid artery stenosis. Furthermore a significant time saving in comparison to standard image interpretation is achievable.

Offshore wind transport and installation vessel

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

The initial objective of the project was to complete a feasibility study to determine the viability of an innovative transportation vessel to be deployed in the installation of offshore wind farms. This included the feasibility of providing a stable-working platform that can be used in harsh offshore environments. A study of current installation contractors and their installation equipment was used to provide a preliminary specification for the installation vessel. A typical barge was selected and a number of hydrodynamic analyses were carried out in order to establish it's on course and operational stability. The analysis proved the stability of the vessel during operation was critical and that in order to utilise the crane's full potential a stabilisation system must be employed. The main aim of the work to date was to establish whether it was feasible to use a stabilisation system on the installation vessel. The spud leg FEED study established that it was feasible to use spud legs to stabilise the vessel. In order to achieve the degree of stability required it is necessary to lift the vessel completely out of the water. This was not the original aim of the study but due to the external loads on the hull it was the only viable option. Lifting the vessel out of the water results in the legs and leg casings becoming very large. This has a number of consequences for the final design. Due to large loads on the legs spud cans must be used to avoid bottom penetration, the spud cans increase the draft of the vessel by 2m. The large loads require larger winches and more reeving to be used, this results in larger pumps and motors, all of which have to be housed. The stabilisation system has been proved to be feasible for a large installation vessel, the cost and physical size are however more excessive than first anticipated. (Author)

Platelet-derived growth factor receptor-β, carrying the activating mutation D849N, accelerates the establishment of B16 melanoma

International Nuclear Information System (INIS)

Suzuki, Shioto; Heldin, Carl-Henrik; Heuchel, Rainer Lothar

2007-01-01

Platelet-derived growth factor (PDGF)-BB and PDGF receptor (PDGFR)-β are mainly expressed in the developing vasculature, where PDGF-BB is produced by endothelial cells and PDGFR-β is expressed by mural cells, including pericytes. PDGF-BB is produced by most types of solid tumors, and PDGF receptor signaling participates in various processes, including autocrine stimulation of tumor cell growth, recruitment of tumor stroma fibroblasts, and stimulation of tumor angiogenesis. Furthermore, PDGF-BB-producing tumors are characterized by increased pericyte abundance and accelerated tumor growth. Thus, there is a growing interest in the development of tumor treatment strategies by blocking PDGF/PDGFR function. We have recently generated a mouse model carrying an activated PDGFR-β by replacing the highly conserved aspartic acid residue (D) 849 in the activating loop with asparagine (N). This allowed us to investigate, in an orthotopic tumor model, the role of increased stromal PDGFR-β signaling in tumor-stroma interactions. B16 melanoma cells lacking PDGFR-β expression and either mock-transfected or engineered to express PDGF-BB, were injected alone or in combination with matrigel into mice carrying the activated PDGFR-β (D849N) and into wild type mice. The tumor growth rate was followed and the vessel status of tumors, i.e. total vessel area/tumor, average vessel surface and pericyte density of vessels, was analyzed after resection. Tumors grown in mice carrying an activated PDGFR-β were established earlier than those in wild-type mice. In this early phase, the total vessel area and the average vessel surface were higher in tumors grown in mice carrying the activated PDGFR-β (D849N) compared to wild-type mice, whereas we did not find a significant difference in the number of tumor vessels and the pericyte abundance around tumor vessels between wild type and mutant mice. At later phases of tumor progression, no significant difference in tumor growth rate was

Conceptual design of the handling and storage system for spent target vessel

Energy Technology Data Exchange (ETDEWEB)

Adachi, Junichi; Sasaki, Shinobu; Kaminaga, Masanori; Hino, Ryutaro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-03-01

A conceptual design of a handling and storage system for spent target vessels has been carried out, in order to establish spent target technology for the neutron scattering facility. The spent target vessels must be treated remotely with high reliability and safety, since they are highly activated and contain the poisonous mercury. The system is composed of a target exchange trolley to exchange the target vessel, remote handling equipment such as manipulators, airtight casks for the spent target vessel, storage pits and so on. This report presents the results of conceptual design study on a basic plan, a handling procedure, main devices and their arrangement of a handling and storage system for the spent target vessels. (author)

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.

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.

After-operating properties of nuclear reactor vessel materials of Lenin atomic ice breaker and prospective of reactor vessels radiation life prolongation

International Nuclear Information System (INIS)

Platonov, P.A.; Shtrombakh, Ya.I.; Amaev, A.D.; Krasikov, E.A.; Korolev, Yu.N.; Zabusov, O.O.; Glushakov, G.M.

2001-01-01

A post-operational state of the icebreaker Lenin reactor vessel metal is investigated. It is shown that a base metal of the icebreaker Lenin reactor vessel is of high quality as by an initial value of critical temperature of embrittlement, so by its radiation resistance. The weld metal possesses a sufficient radiation resistance but has an insufficient initial ductile-brittle transition temperature (approximately 63 Deg C). It is necessary to note that the final stage of operation for nuclear steam-generating plant should be carried out at the coolant temperature as high as possible [ru

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)

Students with Attention Deficit Disorder Carrying out Tasks of Reading Comprehension and Text Production: A Comparative Study in 4th-Grade Students of Primary Education in Chile

OpenAIRE

Fabián Andrés Inostroza-Inostroza

2017-01-01

The present article aims to compare the performance in students with Attention Deficit Disorder to those who do not present it, in tasks of reading comprehension and text production carried out by students attending the fourth grade of primary education. This quantitative, non-experimental comparative study aims to provide evidence regarding the way in which this condition limits the learning outcomes in the tasks of comprehension and production of texts, language, and communication. One the ...

Development of improved SGV480 steel plate for containment vessel in PWR plants

Energy Technology Data Exchange (ETDEWEB)

Murakami, Norioki [Advanced Nuclear Equipment Research Inst., Tokyo (Japan); Morikage, Yasushi; Okayama, Yutaka; Higashikubo, Tomohiro

2001-01-01

When a nuclear containment vessel made of steel plate at PWR plants in Japan is produced, SGV480 steel plate made by annealing method according to JIS G3118 is usually used in main. And, when thickness of welding portion of the vessel is larger than 38 mm, as heat treatment after welding is regulated to carry out according to the ministerial ordinance, it is difficult in actual to carry out the heat treatment of the actual welded portions. In a leading plant, approval of welding using a special method without heat treatment less than 47.25 mm of SGV480 carbon steel plate for JIS G3118 middle and ordinary pressure vessel was carried out to supply it for actual use. And, it is required for protection of welding fracture to carry out pre-heat treatment before welding. Because of increasing plate thickness requiring for lower temperature and more seismic resistance in construction condition, in order to produce a containment vessel without heat treatment after welding, more toughness is required for using material and welded portion. Therefore, a new SGV480 steel plate was developed by using TMCP method of modern steel manufacturing technology, to establish lower carbon equivalence and finer texture with upgrading of both toughness and weldability, without heat treatment after welding and pre-heat treatment before welding, at the Shin-Nippon Steel Co, Ltd. and Kawasaki Steel, Co. Ltd., respectively. (G.K.)

Inservice inspection of Halden BWR pressure vessel

International Nuclear Information System (INIS)

Foerli, O.; Hernes, T.

1978-01-01

A description is given of how the recertification inspection of the 20 years old Halden Reactor pressure vessel was carried out in accordance with the latest ASME-CODES, despite the fact that inspection accessibility was poor. As no volumetric inspection had been carried out since the preservice radiography in 1957, the ultrasonic inspection included the high flux region of all welds. In total 70% of longitudinal welds and 20% of bottom circumferential welds were inspected as well as the bottom nozzle connection. The vessel was not designed with provisions for inservice inspection, the welds are unaccessible from the outside and removal of the lid is virtually impossible. The ultrasonic probes could only be loaded through 77 mm diameter holes in the top lid and remotely positioned inside the vessel. The inspection was performed using 450C and 60OC 1 MHz angle probes and 2.25 MHz normal probes in immersion technique. In a zone around the welds, small regions with lack of bonding between the stainless steel cladding and the boiler steel were revealed. One root defect known and accepted from the preservice radiographs was examined. The defect was found to be 6x30mm as a maximum and well within acceptable limits according to the fracture mechanics analysis method recommended in ASME X1. The inspection required a period of three weeks' work in the reactor hall. (UK)

Feasibility of and methodology for thermal annealing an embrittled reactor vessel. Volume 1. Program overview. Final report

International Nuclear Information System (INIS)

Mager, T.R.

1983-01-01

An EPRI sponsored program was carried out by Westinghouse to determine the extent of fracture toughness recovery as a function of annealing time and temperature for neutron embrittlement sensitive reactor vessel material and to develop an optimal thermal anneal procedure for field applications. Program materials were three weldments fabricated by Combustion Engineering, Inc., from the same heat of A533 Grade B Class 1 plate material and the same heat of MnMoNi weld wire. The only variables were the target copper level and the welding flux which was Linde Grade 80 and Linde 0091. Weldments of 0.22, 0.36, and 0.41 wt % copper were produced. It was concluded from this study that excellent recovery of all properties could be achieved by annealing at 850 0 F (454 0 C) and above for 168 hours. Such an annealing resulted in ductile-brittle transition temperature shift recovery of 80 to 100%, and reirradiation after this annealing indicated that the ductile-brittle transition temperature shift appears to continue at the rate which would have been expected had no anneal been performed. System limitations were identified for both wet and dry annealing methods

Electromagnetic forces on a metallic Tokamak vacuum vessel following a disruptive instability

International Nuclear Information System (INIS)

Eckhartt, D.

1979-04-01

During a 'hard' disruptive instability of a Tokamak plasma the current-carrying plasma is lost within a very short time, typically few milliseconds. If the plasma is contained in a metallic vacuum vessel, electric currents are set up in the vessel following the disappearance of the plasma current. These vessel currents together with the magnetic fields intersecting the vessel generate electromagnetic forces which appear as mechanical loads on the vessel. In the following note it is assumed that the vacuum vessel is surrounded by an 'outer equivalent' or 'flux-conserving' shell having a characteristic time of magnetic field penetration which is long compared to the time of existence of the vessel currents. This property defines the distribution of vessel current densities (and hence the load distribution) without referring to the exact mechanism or time sequence of events by which the plasma current is lost. Numerical examples of the electromagnetic force distribution from this model refer to parameters of the JET-device with the simplifying assumption of circular cross-sections for plasma current, vacuum vessel, and outer equivalent shell. (orig.)

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

International Nuclear Information System (INIS)

Silverblatt, B.L.

1977-01-01

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

New baking system for the RFX vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Collarin, P.; Luchetta, A.; Sonato, P.; Toigo, V.; Zaccaria, P.; Zollino, G. [Universita di Padova (Italy)

1996-12-31

A heating system based on eddy currents has been developed for the vacuum vessel of the RFX Reversed Field Pinch device. After a testing phase, carried out at low power, the final power supply system has been designed and installed. It has been used during last year to bake out the vessel and the graphite first wall up to 320{degree}C. Recently the heating system has been completed with a control system that allows for baking sessions with an automatic control of the vacuum vessel temperature and for pulse sessions with a heated first wall. After the description of the preliminary analyses and tests, and of the main characteristics of the power supply and control systems, the experimental results of the baking sessions performed during last year are presented. 6 refs., 7 figs.

New baking system for the RFX vacuum vessel

International Nuclear Information System (INIS)

Collarin, P.; Luchetta, A.; Sonato, P.; Toigo, V.; Zaccaria, P.; Zollino, G.

1996-01-01

A heating system based on eddy currents has been developed for the vacuum vessel of the RFX Reversed Field Pinch device. After a testing phase, carried out at low power, the final power supply system has been designed and installed. It has been used during last year to bake out the vessel and the graphite first wall up to 320 degree C. Recently the heating system has been completed with a control system that allows for baking sessions with an automatic control of the vacuum vessel temperature and for pulse sessions with a heated first wall. After the description of the preliminary analyses and tests, and of the main characteristics of the power supply and control systems, the experimental results of the baking sessions performed during last year are presented. 6 refs., 7 figs

Design of High Temperature Reactor Vessel Using ANSYS Software

International Nuclear Information System (INIS)

Bandriyana; Kasmudin

2003-01-01

Design calculation and evaluation of material strength for high temperature reactor vessel based on the design of HTR-10 high temperature reactor vessel were carried out by using the ANSYS 5.4 software. ANSYS software was applied to calculate the combined load from thermal and pressure load. Evaluation of material strength was performed by calculate and determine the distribution of temperature, stress and strain in the thickness direction of vessel, and compared with its material strength for designed. The calculation was based on the inner wall temperature of vessel of 600 o C and the outer temperature of 500 and 600 o C. Result of calculation gave the maximum stress for outer temperature of 600 o C was 288 N/ mm 2 and strain of 0.000187. For outer temperature of 500 o C the maximum stress was 576 N/ mm 2 and strain of 0.003. Based on the analysis result, the material of steel SA 516-70 with limited stress for design of 308 N/ mm 2 can be used for vessel material with outer wall temperature of 600 o C

Radon diffusion in polymer vessels using CR-39 solid state nuclear track detector

Energy Technology Data Exchange (ETDEWEB)

Carneiro, Andre Cavalcanti; Menezes, Maria Angela de B.C.; Rocha, Zildete; Pereira, Marcio Tadeu, E-mail: andreccarneiro@gmail.com, E-mail: menezes@cdtn.br, E-mail: zildete@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Santos, Talita de Oliveira; Lara, Evelise Gomes; Braga, Mario Roberto Martins S.S., E-mail: mariomartins@gmail.com, E-mail: evelise.lara@gmail.com, E-mail: talitaolsantos@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

2015-07-01

At CDTN/CNEN, the method to determine {sup 226}Ra in several matrices by gamma spectrometry is already established; however, the method should be improved. This paper is about the first step of this improvement. Several polymer vessels were studied verifying the effect of radiolysis on the walls of the vessel. A test about the diffusion of {sup 222}Rn through the walls was carried out using the CR-39 solid state nuclear track detector. The results pointed out that the vessel made up by acrylic material is the best candidate to replace the vessel actually used. (author)

Detection of solvent losses (entrainment) in gas streams of process vessels using radioisotope tracing techniques

International Nuclear Information System (INIS)

Wan Zakaria Wan Muhamad Tahir; Juhari Mohd Yusof

2002-01-01

Liquid droplets (MDEA aqueous solution) entrained in the gas streams can cause severe problems on chemical plants. On-line detection of liquid entrainment (carry over) into gas streams from process vessel is investigated using radioisotope iodine ( 131 I). In order to obtain information on whether there is any carry-over of MDEA in the vapour space leaving from the process system, a number of test and calibration injections involving the released of certain amount of tracer activity (mCi) at the inlet and overhead lines of the process vessels were made using a special injection device. MDEA solvent- tagged tracer in the overhead line of the designated process vessels was monitored using radiation scintillation detectors mounted externally at specified locations of the vessels. Output pulses (response curves) with respect to time of measurements from all detectors were plotted and analysed for the finger prints of solvent losses leaving the vessels. From this study, no distinguishable peaks were detected at the outlet vessels of the overhead lines. Thus, no significant MDEA solvent losses in the form of vapour being discovered along the gas streams due to the process taking place in the system. (Author)

Manipulator for testing a top-opened reactor pressure vessel

International Nuclear Information System (INIS)

Bauer, R.; Kastl, H.

1991-01-01

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

Automated grading system for evaluation of ocular redness associated with dry eye

Directory of Open Access Journals (Sweden)

Rodriguez JD

2013-06-01

Full Text Available John D Rodriguez,1 Patrick R Johnston,1 George W Ousler III,1 Lisa M Smith,1 Mark B Abelson1,21Ora, Inc, Andover, MA, USA; 2Department of Ophthalmology, Harvard Medical School, Boston, MA, USABackground: We have observed that dry eye redness is characterized by a prominence of fine horizontal conjunctival vessels in the exposed ocular surface of the interpalpebral fissure, and have incorporated this feature into the grading of redness in clinical studies of dry eye.Aim: To develop an automated method of grading dry eye-associated ocular redness in order to expand on the clinical grading system currently used.Methods: Ninety nine images from 26 dry eye subjects were evaluated by five graders using a 0–4 (in 0.5 increments dry eye redness (Ora CalibraTM Dry Eye Redness Scale [OCDER] scale. For the automated method, the Opencv computer vision library was used to develop software for calculating redness and horizontal conjunctival vessels (noted as "horizontality". From original photograph, the region of interest (ROI was selected manually using the open source ImageJ software. Total average redness intensity (Com-Red was calculated as a single channel 8-bit image as R − 0.83G − 0.17B, where R, G and B were the respective intensities of the red, green and blue channels. The location of vessels was detected by normalizing the blue channel and selecting pixels with an intensity of less than 97% of the mean. The horizontal component (Com-Hor was calculated by the first order Sobel derivative in the vertical direction and the score was calculated as the average blue channel image intensity of this vertical derivative. Pearson correlation coefficients, accuracy and concordance correlation coefficients (CCC were calculated after regression and standardized regression of the dataset.Results: The agreement (both Pearson's and CCC among investigators using the OCDER scale was 0.67, while the agreement of investigator to computer was 0.76. A multiple

Safety of steel vessel Magnox pressure circuits

International Nuclear Information System (INIS)

Stokoe, T.Y.; Bolton, C.J.; Heffer, P.J.H.

1991-01-01

The maintenance of pressure circuit integrity is fundamental to nuclear safety at the steel vessel Magnox stations. To confirm continued pressure circuit integrity the CEGB, as part of the Long Term Safety Review, has carried out extensive assessment and inspection in recent years. The assessment methods and inspection techniques employed are based on the most modern available. Reactor pressure vessel integrity is confirmed by a combination of arguments including safety factors inferred from the successful pre-service overpressure test, leak-before-break analysis and probabilistic assessment. In the case of other parts of the pressure circuits that are more accessible, comprising the boiler shells and interconnecting gas duct work, in-service inspection is a major element of the safety substantiation. The assessment and inspection techniques and the materials property data have been underpinned for many years by extensive research and development programmes and in-reactor monitoring of representative samples has also been undertaken. The paper summarises the work carried out to demonstrate the long term integrity of the Magnox pressure circuits and provides examples of the results obtained. (author)

Stromal laminin chain distribution in normal, hyperplastic and malignant oral mucosa: relation to myofibroblast occurrence and vessel formation.

Science.gov (United States)

Franz, Marcus; Wolheim, Anke; Richter, Petra; Umbreit, Claudia; Dahse, Regine; Driemel, Oliver; Hyckel, Peter; Virtanen, Ismo; Kosmehl, Hartwig; Berndt, Alexander

2010-04-01

The contribution of stromal laminin chain expression to malignant potential, tumour stroma reorganization and vessel formation in oral squamous cell carcinoma (OSCC) is not fully understood. Therefore, the expression of the laminin chains alpha2, alpha3, alpha4, alpha5 and gamma2 in the stromal compartment/vascular structures in OSCC was analysed. Frozen tissue of OSCC (9x G1, 24x G2, 8x G3) and normal (2x)/hyperplastic (11x) oral mucosa was subjected to laminin chain and alpha-smooth muscle actin (ASMA) immunohistochemistry. Results were correlated to tumour grade. The relation of laminin chain positive vessels to total vessel number was assessed by immunofluorescence double labelling with CD31. Stromal laminin alpha2 chain significantly decreases and alpha3, alpha4, alpha5 and gamma2 chains and also ASMA significantly increase with rising grade. The amount of stromal alpha3, alpha4 and gamma2 chains significantly increased with rising ASMA positivity. There is a significant decrease in alpha3 chain positive vessels with neoplastic transformation. Mediated by myofibroblasts, OSCC development is associated with a stromal up-regulation of laminin isoforms possibly contributing to a migration promoting microenvironment. A vascular basement membrane reorganization concerning alpha3 and gamma2 chain laminins during tumour angioneogenesis is suggested.

Cylindrical prestressed concrete pressure vessel for a nuclear power plant

International Nuclear Information System (INIS)

Horner, M.; Hodzic, A.; Haferkamp, D.

1976-01-01

A prestressed concrete pressure vessel for a HTGR is proposed which encloses, in addition to the reactor core, not only the heat-exchanging facilities but also the turbine unit. The reinforcement of the cylindrical concrete body is to be carried out with special care, it is provided for horizontal tendons, the prestressed concrete pressure vessel has a wire-winding device, while the longitudinal reinforcement is achieved by tendous guided in parallel to the vesses axes through the interspaces between the pods. (UWI) [de

Minimum critical crack depths in pressure vessels guidelines for nondestructive testing

International Nuclear Information System (INIS)

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

1983-09-01

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

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.

Onshore preparedness for hazardous chemical marine vessel accidents: A case study

Directory of Open Access Journals (Sweden)

Faisel T. Illiyas

2016-09-01

Full Text Available Hazardous and noxious substances (HNS are widely transported in marine vessels to reach every part of the world. Bulk transportation of hazardous chemicals is carried out in tank container–carrying cargo ships or in designed vessels. Ensuring the safety of HNS containers during maritime transportation is critically important as the accidental release of any substance may be lethal to the on-board crew and marine environment. A general assumption in maritime accidents in open ocean is that it will not create any danger to the coastal population. The case study discussed in this article throws light on the dangers latent in maritime HNS accidents. An accident involving an HNS-carrying marine vessel in the Arabian Sea near the coast of Yemen became a safety issue to the coastal people of Kasargod District of Kerala, India. The ship carried more than 4000 containers, which were lost to the sea in the accident. Six HNS tank containers were carried by the waves and shored at the populated coast of Kasargod, more than 650 nautical miles east from the accident spot. The unanticipated sighting of tank containers in the coast and the response of the administration to the incident, the hurdles faced by the district administration in handling the case, the need for engaging national agencies and lessons learned from the incident are discussed in the article. This case study has proven that accidents in the open ocean have the potential to put the coastal areas at risk if the on-board cargo contains hazardous chemicals. Littoral nations, especially those close to the international waterlines, must include hazardous chemical spills to their oil spill contingency plans.

76 FR 67253 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel THE GIFT; Invitation for...

Science.gov (United States)

2011-10-31

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD 2011 0131] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel THE GIFT; Invitation for Public Comments AGENCY... GIFT is: INTENDED COMMERCIAL USE OF VESSEL: ``Passenger carrying.'' GEOGRAPHIC REGION: ``ME, NH, MA, RI...

The value of [18F]FDG-PET in the diagnosis of large-vessel vasculitis and the assessment of activity and extent of disease

International Nuclear Information System (INIS)

Walter, Martin A.; Mueller-Brand, Jan; Nitzsche, Egbert U.; Melzer, Ralph A.; Tyndall, Alan; Schindler, Christian

2005-01-01

This study was performed to investigate the value of 18 F-fluorodeoxyglucose positron emission tomography ([ 18 F]FDG-PET) in the diagnosis of large-vessel vasculitis and the assessment of activity and extent of disease. Twenty-six consecutive patients (21 females, 5 males; median age - years, range 17-86 years) with giant cell arteritis or Takayasu's arteritis were examined with [ 18 F]FDG-PET. Follow-up scans were performed in four patients. Twenty-six age- and gender-matched controls (21 females, 5 males; median age 71 years, range 17-86 years) were included. The severity of large-vessel [ 18 F]FDG uptake was visually graded using a four-point scale. C-reactive protein (CRP) and the erythrocyte sedimentation rate (ESR) were measured and correlated with [ 18 F]FDG-PET results by logistic regression. [ 18 F]FDG-PET revealed pathological findings in 18 of 26 patients. Three scans were categorised as grade I, 12 as grade II and 3 as grade III arteritis. Visual grade was significantly correlated with both CRP and ESR levels (p=0.002 and 0.007 respectively; grade I: CRP 4.0 mg/l, ESR 6 mm/h; grade II: CRP 37 mg/l, ESR 46 mm/h; grade III: CRP 172 mg/l, ESR 90 mm/h). Overall sensitivity was 60% (95% CI 40.6-77.3%), specificity 99.8% (95% CI 89.1-100%), positive predictive value 99.7% (95% CI 77-100%), negative predictive value 67.9% (95% CI 49.8-80.9%) and accuracy 78.6% (95% CI 65.6-88.4%). In patients presenting with a CRP 18 F]FDG-PET is highly effective in assessing the activity and the extent of large-vessel vasculitis. Visual grading was validated as representing the severity of inflammation. Its use is simple and provides high specificity, while high sensitivity is achieved by scanning in the state of active inflammation. (orig.)

Effect of passing vessels on a moored ship

Energy Technology Data Exchange (ETDEWEB)

Lean, G H; Price, W A

1977-11-01

The effect of passing vessels on a moored ship was investigated by a series of model tests carried out at the Hydraulics Research Station for the Esso Petroleum Co. Ltd., transportation department in connection with their oil jetty at Milford Haven. A main conclusion was that the forces appeared to be due to the pressure gradients associated with the pattern of flow that accompanies the passing ship rather than with the wave system. Slack lines are to be avoided, and some relief in maximum line loads can be achieved by increasing the pretension. The results included the effects of passing vessel speed and ship clearance and draft.

77 FR 44475 - Security Zones; Seattle's Seafair Fleet Week Moving Vessels, Puget Sound, WA

Science.gov (United States)

2012-07-30

...-AA87 Security Zones; Seattle's Seafair Fleet Week Moving Vessels, Puget Sound, WA AGENCY: Coast Guard... temporary rule, call or email Lieutenant Junior Grade Anthony P. LaBoy, Sector Puget Sound, Waterways Management Division, U.S. Coast Guard; telephone 206-217-6323, email SectorPugetSound[email protected] . If you...

Mooring system for a permanently moored storage vessel at an offshore site

Energy Technology Data Exchange (ETDEWEB)

Flory, J.F.

1983-01-24

A vessel, e.g. a storage vessel, is permanently moored by means such as a yoke pivoted on the forecastle of the vessel to a mooring leg, e.g. a riser or anchor chain, which is attached to a base located on the ocean floor. Mounted on the vessel are tension, exerting means e.g. counterweights, springs, winches, etc., operably connected with the mooring leg for applying tension e.g. by lifting the yoke. The top of the mooring leg is connected to the end of the yoke through a mooring swivel and gimbaled mooring table or a universal joint. A fluid swivel may be located above the mooring table or about a load-carrying shaft connected to the mooring leg.

78 FR 30961 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel LITTLE DUTCH; Invitation for...

Science.gov (United States)

2013-05-23

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD-2013-0057] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel LITTLE DUTCH; Invitation for Public Comments AGENCY... LITTLE DUTCH is: Intended Commercial Use of Vessel: ``Carrying up to six passengers for day trips, sunset...

Distribution of Vascular Patterns in Different Subtypes of Renal Cell Carcinoma. A Morphometric Study in Two Distinct Types of Blood Vessels.

Science.gov (United States)

Ruiz-Saurí, Amparo; García-Bustos, V; Granero, E; Cuesta, S; Sales, M A; Marcos, V; Llombart-Bosch, A

2017-07-01

To analyze the presence of mature and immature vessels as a prognostic factor in patients with renal cell carcinoma and propose a classification of renal cancer tumor blood vessels according to morphometric parameters. Tissue samples were obtained from 121 renal cell carcinoma patients who underwent radical nephrectomy. Staining with CD31 and CD34 was used to differentiate between immature (CD31+) and mature (CD34+) blood vessels. We quantified the microvascular density, microvascular area and different morphometric parameters: maximum diameter, minimum diameter, major axis, minor axis, perimeter, radius ratio and roundness. We found that the microvascular density was higher in CD31+ than CD34+ vessels, but CD34+ vessels were larger than CD31+ vessels, as well as being strongly correlated with the ISUP tumor grade. We also identified four vascular patterns: pseudoacinar, fascicular, reticular and diffuse. Pseudoacinar and fascicular patterns were more frequent in clear cell renal cell carcinoma (37.62 and 35.64% respectively), followed by reticular pattern (21.78%), while in chromophobe tumors the reticular pattern predominated (90%). The isolated pattern was present in all papillary tumors (100%). In healthy renal tissue, the pseudoacinar and isolated patterns were differentially found in the renal cortex and medulla respectively. We defined four distinct vascular patterns significantly related with the ISUP tumor grade in renal cell carcinomas. Further studies in larger series are needed in order to validate these results. Analysis of both mature and immature vessels (CD34+ and CD31+) provides additional information when evaluating microvascular density.

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

International Nuclear Information System (INIS)

Stahlberg, R.

1978-01-01

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

A prototype knowledge based system for pressure vessel design

Energy Technology Data Exchange (ETDEWEB)

Gunnarsson, L.

1991-11-22

The usage of expert system techniques in the area of mechanical engineering design has been studied. A prototype expert system for pressure vessel design has been developed. The work has been carried out in two steps. Firstly, a pre-processor for the finite element system PCFEMP, named INFEMP, was developed. Secondly, an expert supported system for pressure vessel design, named PVES, was developed. Both INFEMP and PVES are integrated to the AutoCAD system, and AutoCAD`s language AutoLISP has been used. A practical example has been investigated to demonstrate the principal ideas of the prototype. (au).

A prototype knowledge based system for pressure vessel design

Energy Technology Data Exchange (ETDEWEB)

Gunnarsson, L.

1991-11-22

The usage of expert system techniques in the area of mechanical engineering design has been studied. A prototype expert system for pressure vessel design has been developed. The work has been carried out in two steps. Firstly, a pre-processor for the finite element system PCFEMP, named INFEMP, was developed. Secondly, an expert supported system for pressure vessel design, named PVES, was developed. Both INFEMP and PVES are integrated to the AutoCAD system, and AutoCAD's language AutoLISP has been used. A practical example has been investigated to demonstrate the principal ideas of the prototype. (au).

A prototype knowledge based system for pressure vessel design

International Nuclear Information System (INIS)

Gunnarsson, L.

1991-01-01

The usage of expert system techniques in the area of mechanical engineering design has been studied. A prototype expert system for pressure vessel design has been developed. The work has been carried out in two steps. Firstly, a pre-processor for the finite element system PCFEMP, named INFEMP, was developed. Secondly, an expert supported system for pressure vessel design, named PVES, was developed. Both INFEMP and PVES are integrated to the AutoCAD system, and AutoCAD's language AutoLISP has been used. A practical example has been investigated to demonstrate the principal ideas of the prototype. (au)

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

Light-water-reactor pressure-vessel surveillance dosimetry using solid-state track recorders

International Nuclear Information System (INIS)

Ruddy, F.H.; Roberts, J.H.; Gold, R.; Preston, C.C.

1983-07-01

The accumulation of neutron dose by the pressure vessel of an operating nuclear power plant results in damage in the form of steel embrittlement. In order to ascertain the safe operating lifetime of the reactor pressure vessel, dosimetric measurements must be made to evaluate the neutron dose to the pressure vessel and relate this dose to the cumulative radiation damage. Advanced dosimetry techniques are being evaluated for surveillance of operating reactors. Solid-state track recorder (SSTR) techniques are included among these advanced dosimetry techniques. Described herein are low neutron fluence calibration and standardization measurements that are being carried out in pressure vessel mockup benchmark neutron fields in the USA, Belgium, and England. In addition, high fluence SSTR dosimetry capsules have been irradiated with metallurgical specimens in a pressure vessel mockup facility. The design and deployment of advances SSTR dosimetry capsules in operating power reactors are also described

CT angiography of intracranial arterial vessels: impact of tube voltage and contrast media concentration on image quality

International Nuclear Information System (INIS)

Ramgren, Birgitta; Holtaas, Stig; Siemund, Roger; Dept. of Radiology, Lund Univ., Lund

2012-01-01

Background Computed tomography angiography (CTA) of intracranial arteries has high demands on image quality. Important parameters influencing vessel enhancement are injection rate, concentration of contrast media and tube voltage. Purpose To evaluate the impact of an increase of contrast media concentration from 300 to 400 mg iodine/mL (mgI/mL) and the effect of a decrease of tube voltage from 120 to 90 kVp on vessel attenuation and image quality in CT angiography of intracranial arteries. Material and Methods Sixty-three patients were included into three protocol groups: Group I, 300 mgI/mL 120 kVp; Group II, 400 mgI/mL 120 kVp; Group III, 400 mgI/mL 90 kVp. Hounsfield units (HU) were measured in the internal carotid artery (ICA) and the M1 and M2 segments of the middle cerebral artery. Image quality grading was performed regarding M1 and M2 segments, volume rendering and general image impression. Results The difference in mean HU in ICA concerning the effect of contrast media concentration was statistically significant (P = 0.03) in favor of higher concentration. The difference in ICA enhancement due to the effect of tube voltage was statistically significant (P < 0.01) in favor of lower tube voltage. The increase of contrast medium concentration raised the mean enhancement in ICA with 18% and the decrease of tube voltage raised the mean enhancement with 37%. Image quality grading showed a trend towards improved grading for higher contrast concentration and lower tube voltage. Statistically significant better grading was found for the combined effect of both measures except for general impression (P 0.01-0.05). Conclusion The uses of highly concentrated contrast media and low tube voltage are easily performed measures to improve image quality in CTA of intracranial vessel

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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.

Fourier series analysis of a cylindrical pressure vessel subjected to axial end load and external pressure

International Nuclear Information System (INIS)

Brar, Gurinder Singh; Hari, Yogeshwar; Williams, Dennis K.

2013-01-01

This paper presents the comparison of a reliability technique that employs a Fourier series representation of random axisymmetric and asymmetric imperfections in a cylindrical pressure vessel subjected to an axial end load and external pressure, with evaluations prescribed by the ASME Boiler and Pressure Vessel Code, Section VIII, Division 2 Rules. The ultimate goal of the reliability technique described herein is to predict the critical buckling load associated with the subject cylindrical pressure vessel. Initial geometric imperfections are shown to have a significant effect on the calculated load carrying capacity of the vessel. Fourier decomposition was employed to interpret imperfections as structural features that can be easily related to various other types of defined imperfections. The initial functional description of the imperfections consists of an axisymmetric portion and a deviant portion, which are availed in the form of a double Fourier series. Fifty simulated shells generated by the Monte Carlo technique are employed in the final prediction of the critical buckling load. The representation of initial geometrical imperfections in the cylindrical pressure vessel requires the determination of respective Fourier coefficients. Multi-mode analyses are expanded to evaluate a large number of potential buckling modes for both predefined geometries in combination with asymmetric imperfections as a function of position within the given cylindrical shell. The probability of the ultimate buckling stress exceeding a predefined threshold stress is also calculated. The method and results described herein are in stark contrast to the “knockdown factor” approach as applied to compressive stress evaluations currently utilized in industry. Further effort is needed to improve on the current design rules regarding column buckling of large diameter pressure vessels subjected to an axial end load and external pressure designed in accordance with ASME Boiler and

Intra-arterial DSA of the mesenterico-spleno-portal vessels

Energy Technology Data Exchange (ETDEWEB)

Busch, H P; Hoevels, J; Prager, P; Strauss, L

1985-01-01

The article examines the application of i.a. DSA for the visualization of mesenterico-spleno-portal veins. Indications are portal hypertension and resectability assessment in pancreas tumours. Compared with conventional angiography, i.a. DSA yields a better demonstration of the splanchnic veins in about 50% of the cases. Advantages of i.a. DSA involve good-quality vessel visualization along with a reduction of examination time and cost. Its disadvantages are low-grade local resolution and strong dependence of picture quality on the patients' cooperation.

The characteristics of the prestressed concrete reactor vessel of the HHT demonstration plant

International Nuclear Information System (INIS)

Schoening, J.; Schwiers, H.G.

1979-01-01

The paper concentrates on the design studies of the HTGR prestressed concrete reactor vessel (PCRV) for the HHT Demonstration Plant. The multi-cavity reactor pressure vessel accommodates all components carrying primary gas, including heat exchangers and gas turbine. For reasons of economics and availability of the reactor plant, generic requirements are made for the PCRV. A short description of the power plant is also presented

Decision support system for the detection and grading of hard exudates from color fundus photographs

Science.gov (United States)

Jaafar, Hussain F.; Nandi, Asoke K.; Al-Nuaimy, Waleed

2011-11-01

Diabetic retinopathy is a major cause of blindness, and its earliest signs include damage to the blood vessels and the formation of lesions in the retina. Automated detection and grading of hard exudates from the color fundus image is a critical step in the automated screening system for diabetic retinopathy. We propose novel methods for the detection and grading of hard exudates and the main retinal structures. For exudate detection, a novel approach based on coarse-to-fine strategy and a new image-splitting method are proposed with overall sensitivity of 93.2% and positive predictive value of 83.7% at the pixel level. The average sensitivity of the blood vessel detection is 85%, and the success rate of fovea localization is 100%. For exudate grading, a polar fovea coordinate system is adopted in accordance with medical criteria. Because of its competitive performance and ability to deal efficiently with images of variable quality, the proposed technique offers promising and efficient performance as part of an automated screening system for diabetic retinopathy.

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.

Assessment of the effects of neutron fluence on Swedish nuclear pressure vessels

International Nuclear Information System (INIS)

Rao, S.

1980-11-01

Nuclear pressure vessels are subject to neutron irradiation during service causing embrittlement. This is one important factor in the overall problem of reactor vessel integrity. At present the irradiation effects are mainly assessed by the Charpy V-notch test. Two measures of embrittlement are defined: the increase of the ductile/brittle transition temperature and the decrease in the upper-shelf energy. The object of the present work is to assess these changes for the Swedish nuclear pressure vessels. On the basis of data from irradiations carried out in other countries and Swedish surveillance programmes, the expected end of life embrittlement is estimated for Swedish vessels. The results show that the embrittlement of most reactor vessels is expected to be quite small. Oskarshamn 1 and PWR-vessels, however, will probably show moderate changes, the former due to the higher copper content, and the latter due to the high end of life fluences. Some of the vessel materials which exhibit marginal properties in the upper-shelf energy, as measured by the Charpy V-notch impact test, are identified. It is recommended that fracture mechanics analyses be applied in these cases. (author)

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.

Ultimate load design and testing of a cylindrical prestressed concrete vessel

International Nuclear Information System (INIS)

Stefanou, G.D.

1982-01-01

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

Failure probability analysis on mercury target vessel

International Nuclear Information System (INIS)

Ishikura, Syuichi; Futakawa, Masatoshi; Kogawa, Hiroyuki; Sato, Hiroshi; Haga, Katsuhiro; Ikeda, Yujiro

2005-03-01

Failure probability analysis was carried out to estimate the lifetime of the mercury target which will be installed into the JSNS (Japan spallation neutron source) in J-PARC (Japan Proton Accelerator Research Complex). The lifetime was estimated as taking loading condition and materials degradation into account. Considered loads imposed on the target vessel were the static stresses due to thermal expansion and static pre-pressure on He-gas and mercury and the dynamic stresses due to the thermally shocked pressure waves generated repeatedly at 25 Hz. Materials used in target vessel will be degraded by the fatigue, neutron and proton irradiation, mercury immersion and pitting damages, etc. The imposed stresses were evaluated through static and dynamic structural analyses. The material-degradations were deduced based on published experimental data. As a result, it was quantitatively confirmed that the failure probability for the lifetime expected in the design is very much lower, 10 -11 in the safety hull, meaning that it will be hardly failed during the design lifetime. On the other hand, the beam window of mercury vessel suffered with high-pressure waves exhibits the failure probability of 12%. It was concluded, therefore, that the leaked mercury from the failed area at the beam window is adequately kept in the space between the safety hull and the mercury vessel by using mercury-leakage sensors. (author)

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.

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

Intrinsic subtypes and tumor grades in breast cancer are associated with distinct 3-D power Doppler sonographic vascular features

International Nuclear Information System (INIS)

Chang, Yeun-Chung; Huang, Yao-Sian; Huang, Chiun-Sheng; Chen, Jeon-Hor; Chang, Ruey-Feng

2014-01-01

Purpose: This study aimed to investigate the three-dimensional (3-D) power Doppler ultrasonographic (PDUS) vascular features of breast carcinoma according to intrinsic subtypes, nodal stage, and tumor grade. Materials and methods: Total 115 receiving mastectomy breast carcinomas (mean size, 2.5 cm; range, 0.7–6.5 cm), including 102 invasive ductal carcinomas (IDC), 10 ductal carcinomas in situ (DCIS), and 3 invasive lobular carcinomas (ILC) diagnosed after mastectomy, were used in this retrospective study. Sixty IDC had nodal status and histopathologic tumor grades available for analysis. Vascular features, including number of vascular trees (NV), longest path length (LPL), total vessel length (TVL), number of bifurcations (NB), distance metric (DM), inflection count metric (ICM), vessel diameter (VD), and vessel-to-volume ratio (VVR) were extracted using 3-D thinning method. The Mann–Whitney U test, Student's t-test, one-way ANOVA, and Kruskal–Wallis test were performed as appropriate. Results: There was no significant difference of vascular features among IDC, DCIS and ILC. Except VD, vascular features in luminal type were significantly lower compared to HER2-enriched or triple negative types (p < 0.05). Compared to ER+ (estrogen receptor positive) tumors, all features in ER− (estrogen receptor negative) tumors were significantly higher (p < 0.01). Despite some significantly higher vascular features in high grade IDC compared to low and intermediate grade, there was no significant correlation between vascular features and nodal stages. Conclusion: Differences in 3-D PDUS vascular features among intrinsic types of IDC are attributed to their ER status. Vascular features extracted by 3-D PDUS correlate with tumor grades but not nodal stage in IDC

Intrinsic subtypes and tumor grades in breast cancer are associated with distinct 3-D power Doppler sonographic vascular features

Energy Technology Data Exchange (ETDEWEB)

Chang, Yeun-Chung [Department of Medical Imaging, National Taiwan University Hospital and National Taiwan University College of Medicine, Taipei 10041, Taiwan, ROC (China); Huang, Yao-Sian [Department of Computer Science and Information Engineering, National Taiwan University, Taipei 10617, Taiwan, ROC (China); Huang, Chiun-Sheng [Department of Surgery, National Taiwan University Hospital and National Taiwan University College of Medicine, Taipei 10041, Taiwan, ROC (China); Graduate Institute of Biomedical Electronics and Bioinformatics, National Taiwan University, Taipei 10617, Taiwan, ROC (China); Chen, Jeon-Hor [Center for Functional Onco-Imaging and Department of Radiological Science, University of California Irvine, California, CA 92868 (United States); Department of Radiology, E-Da Hospital and I-Shou University, Kaohsiung 82445, Taiwan, ROC (China); Chang, Ruey-Feng, E-mail: rfchang@csie.ntu.edu.tw [Department of Computer Science and Information Engineering, National Taiwan University, Taipei 10617, Taiwan, ROC (China); Graduate Institute of Biomedical Electronics and Bioinformatics, National Taiwan University, Taipei 10617, Taiwan, ROC (China)

2014-08-15

Purpose: This study aimed to investigate the three-dimensional (3-D) power Doppler ultrasonographic (PDUS) vascular features of breast carcinoma according to intrinsic subtypes, nodal stage, and tumor grade. Materials and methods: Total 115 receiving mastectomy breast carcinomas (mean size, 2.5 cm; range, 0.7–6.5 cm), including 102 invasive ductal carcinomas (IDC), 10 ductal carcinomas in situ (DCIS), and 3 invasive lobular carcinomas (ILC) diagnosed after mastectomy, were used in this retrospective study. Sixty IDC had nodal status and histopathologic tumor grades available for analysis. Vascular features, including number of vascular trees (NV), longest path length (LPL), total vessel length (TVL), number of bifurcations (NB), distance metric (DM), inflection count metric (ICM), vessel diameter (VD), and vessel-to-volume ratio (VVR) were extracted using 3-D thinning method. The Mann–Whitney U test, Student's t-test, one-way ANOVA, and Kruskal–Wallis test were performed as appropriate. Results: There was no significant difference of vascular features among IDC, DCIS and ILC. Except VD, vascular features in luminal type were significantly lower compared to HER2-enriched or triple negative types (p < 0.05). Compared to ER+ (estrogen receptor positive) tumors, all features in ER− (estrogen receptor negative) tumors were significantly higher (p < 0.01). Despite some significantly higher vascular features in high grade IDC compared to low and intermediate grade, there was no significant correlation between vascular features and nodal stages. Conclusion: Differences in 3-D PDUS vascular features among intrinsic types of IDC are attributed to their ER status. Vascular features extracted by 3-D PDUS correlate with tumor grades but not nodal stage in IDC.

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...

Reasons and remedies of inland passenger vessels accidents in Bangladesh

Science.gov (United States)

Rashid, Cdr Kaosar; Islam, Muhammad Rabiul

2017-12-01

The waterways are very important means of communication in Bangladesh. Every year over 95 million passengers are carried through this route. But, this important mode of transport is ridden with tragic disasters every year, incurring a heavy toll of human lives. In last twenty years (1994 to 2014), around 5,500 people have died and 1,500 gone missing in 658 launch disasters. The inland routes of Barisal, Bhola, Chandpur and Patuakhali and their connected water ways to Dhaka and Chittagong are found to be more accident prone. Lack of Awareness, boundless operation of unfit vessels, overloading of passengers, recruitment of unskilled crews, poor capacity of relevant bodies and low standard maintenance of Inland Water Transport (IWT) channels, poor weather forecasting, profit centered attitude of vessel owners and corruption are initiating these deadly accidents. Despite of a number of initiatives by the government, concerned departments and foreign consultants, the safety aspect of the inland passenger vessels still remains in dark. Combined effort of Department of Shipping, BIWTA, and the attitude of vessels owners as well as passengers are very essential in this respect.

Autonomy and Task Performance: Explaining the Impact of Grades on Intrinsic Motivation

Science.gov (United States)

Pulfrey, Caroline; Darnon, Celine; Butera, Fabrizio

2013-01-01

The use of grades to motivate constitutes an unresolved theoretical controversy. In 2 experiments carried out with different age groups and academic tracks, a standard-grade condition was compared with a condition in which differential scoring engendered higher grades and with a no-grade condition. The relative power of task performance and task…

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

International Nuclear Information System (INIS)

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

1977-01-01

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

Testing of Full Scale Flight Qualified Kevlar Composite Overwrapped Pressure Vessels

Science.gov (United States)

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

2007-01-01

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

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

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

Hyperintense vessels on FLAIR: A useful non-invasive method for assessing intracerebral collaterals

International Nuclear Information System (INIS)

Liu Wenhua; Xu Gelin; Yue Xuanye; Wang Xiaoliang; Ma Minmin; Zhang Renliang; Wang Handong; Zhou Changsheng; Liu Xinfeng

2011-01-01

Objective: This study was aimed to evaluate relationship between hyperintense vessels (HV) on fluid-attenuated inversion recovery (FLAIR) and artery steno-occlusion related intracerebral collaterals. Materials and methods: A total of 233 patients with 260 atherosclerotic lesions in the M1 segment of the middle cerebral artery (MCA) were examined with FLAIR and digital subtraction angiography (DSA). HV were graded as 0, 1, 2 and 3 by its distributions in the MCA territory. Grade 0 indicated no HV; Grade 1 indicated the HV limited in Sylvian fissure; Grade 2 indicated the HV limited in Sylvian fissure and the temporal-occipital junction; Grade 3 indicated the HV extended to frontal-parietal lobes. Collateral blood flows were classified by DSA results. The relationship between HV grades and patterns of collateral flows was analyzed. Results: HV were observed in 76 out of 260 hemispheres. For patients with Grade 1 HV, most of their collateral flows (80.8%) were antegrade; for patients with Grade 2, the retrograde leptomeningeal flows were commonly manifested as anterior cerebral artery to MCA (75%); for patients with Grade 3 HV, most of the retrograde leptomeningeal flows were manifested as posterior cerebral artery to MCA (81.8%). As the grade HV increased, the frequency of retrograde leptomeningeal collateral from ACA to MCA decreased (100% to 75% and to 18.2%), and increased (0% to 25% and to 81.8%) for the retrograde leptomeningeal collateral via PCA to MCA (P < 0.001). Conclusions: The HV could assess non-invasively intracerebral collaterals in patients with steno-occlusive lesions of M1 segment of MCA.

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)

The value of [{sup 18}F]FDG-PET in the diagnosis of large-vessel vasculitis and the assessment of activity and extent of disease

Energy Technology Data Exchange (ETDEWEB)

Walter, Martin A.; Mueller-Brand, Jan; Nitzsche, Egbert U. [University Hospital Basel, Institute of Nuclear Medicine, Basel (Switzerland); Melzer, Ralph A.; Tyndall, Alan [University Hospital Basel, Division of Rheumatology (Switzerland); Schindler, Christian [University Hospital Basel, Institute of Social and Preventive Medicine (Switzerland)

2005-06-01

This study was performed to investigate the value of{sup 18}F-fluorodeoxyglucose positron emission tomography ([{sup 18}F]FDG-PET) in the diagnosis of large-vessel vasculitis and the assessment of activity and extent of disease. Twenty-six consecutive patients (21 females, 5 males; median age - years, range 17-86 years) with giant cell arteritis or Takayasu's arteritis were examined with [{sup 18}F]FDG-PET. Follow-up scans were performed in four patients. Twenty-six age- and gender-matched controls (21 females, 5 males; median age 71 years, range 17-86 years) were included. The severity of large-vessel [{sup 18}F]FDG uptake was visually graded using a four-point scale. C-reactive protein (CRP) and the erythrocyte sedimentation rate (ESR) were measured and correlated with [{sup 18}F]FDG-PET results by logistic regression. [{sup 18}F]FDG-PET revealed pathological findings in 18 of 26 patients. Three scans were categorised as grade I, 12 as grade II and 3 as grade III arteritis. Visual grade was significantly correlated with both CRP and ESR levels (p=0.002 and 0.007 respectively; grade I: CRP 4.0 mg/l, ESR 6 mm/h; grade II: CRP 37 mg/l, ESR 46 mm/h; grade III: CRP 172 mg/l, ESR 90 mm/h). Overall sensitivity was 60% (95% CI 40.6-77.3%), specificity 99.8% (95% CI 89.1-100%), positive predictive value 99.7% (95% CI 77-100%), negative predictive value 67.9% (95% CI 49.8-80.9%) and accuracy 78.6% (95% CI 65.6-88.4%). In patients presenting with a CRP <12 mg/l or an ESR <12 mm/h, logistic regression revealed a sensitivity of less than 50%. In patients with high CRP/ESR levels, sensitivity was 95.5%/80.7%. [{sup 18}F]FDG-PET is highly effective in assessing the activity and the extent of large-vessel vasculitis. Visual grading was validated as representing the severity of inflammation. Its use is simple and provides high specificity, while high sensitivity is achieved by scanning in the state of active inflammation. (orig.)

The clinical impact of mean vessel size and solidity in breast carcinoma patients.

Directory of Open Access Journals (Sweden)

Lars Tore Gyland Mikalsen

Full Text Available Angiogenesis quantification, through vessel counting or area estimation in the most vascular part of the tumour, has been found to be of prognostic value across a range of carcinomas, breast cancer included. We have applied computer image analysis to quantify vascular properties pertaining to size, shape and spatial distributions in photographed fields of CD34 stained sections. Aided by a pilot (98 cases, seven parameters were selected and validated on a separate set from 293 breast cancer patients. Two new prognostic markers were identified through continuous cox regression with endpoints breast cancer specific survival and distant disease free survival: The average size of the vessels as measured by their perimeter (p = 0.003 and 0.004, respectively, and the average complexity of the vessel shapes measured by their solidity (p = 0.004 and 0.004. The Hazard ratios for the corresponding median-dichotomized markers were 2.28 (p = 0.005 and 1.89 (p = 0.016 for the mean perimeter and 1.80 (p = 0.041 and 1.55 (p = 0.095 for the shape complexity. The markers were associated with poor histologic type, high grade, necrosis, HR negativity, inflammation, and p53 expression (vessel size only. Both markers were found to strongly influence the prognostic properties of vascular invasion (VI and disseminated tumour cells in the bone marrow. The latter being prognostic only in cases with large vessels (p = 0.004 and 0.043 or low complexity (p = 0.018 and 0.024, but not in the small or complex vessel groups (p>0.47. VI was significant in all groups, but showed greater hazard ratios for small and low complexity vessels (6.54-11.2 versus large and high complexity vessels (2.64-3.06. We find that not only the overall amount of produced vasculature in angiogenic hot-spots is of prognostic significance, but also the morphological appearance of the generated vessels, i.e. the size and shape of vessels in the studied hot spots.

Ductile fracture toughness of modified A 302 grade B plate materials. Volume 2

International Nuclear Information System (INIS)

McCabe, D.E.; Manneschmidt, E.T.; Swain, R.L.

1997-02-01

The objective of this work was to develop ductile fracture toughness data in the form of J-R curves for modified A 302 grade B plate materials typical of those used in fabricating reactor pressure vessels. A previous experimental study at Materials Engineering Associates (MEA) on one particular heat of A 302 grade B plate showed decreasing J-R curves with increased specimen thickness. This characteristic has not been observed in numerous tests made on the more recent production materials of A 533 grade B and A 508 class 2 pressure vessel steels. It was unknown if the departure from norm for the MEA material was a generic characteristic for all heats of A 302 grade B steels or just unique to that one particular plate. Seven heats of modified A 302 grade B steel and one heat of vintage A 533 grade B steel were provided to this project by the General Electric Company of San Jose, California. All plates were tested for chemical content, tensile properties, Charpy transition temperature curves, drop-weight nil-ductility transition (NDT) temperature, and J-R curves. Tensile tests were made in the three principal orientations and at four temperatures, ranging from room temperature to 550 degrees F (288 degrees C). Charpy V-notch transition temperature curves were obtained in longitudinal, transverse, and short transverse orientations. J-R curves were made using four specimen sizes (1/2T, IT, 2T, and 4T). None of the seven heats of modified A 302 grade showed size effects of any consequence on the J-R curve behavior. Crack orientation effects were present, but none were severe enough to be reported as atypical. A test temperature increase from 180 to 550 degrees F (82 to 288 degrees C) produced the usual loss in J-R curve fracture toughness. Generic J-R curves and mathematical curve fits to the same were generated to represent each heat of material. This volume is a compilation of all data developed

Tribology aspects of a pressure vessel closure subjected to pressure cycling

International Nuclear Information System (INIS)

George, A.F.; Williams, M.E.

1988-04-01

A repair method being considered for a steel pressure vessel is to cut away the faulty part leaving an unreinforced circular hole in the curved wall and cover it with a sealed plate placed inside. In order to investigate the structural properties of such a repair a large model vessel (6m by 2m) was tested under pressure (about 2.5 MPa) and pressure cycling. This cycling caused relative movements at the loaded interface between the lid and the vessel. A tribological examination of the rubbing surfaces was carried out. The tribological examination is described and a small supporting programme of laboratory scaling tests. It gives the results and attempts to interpret them with particular attention given to wear, fretting fatigue and scaling to plant conditions. (author)

Development of cold moderator vessel for the spallation neutron source. Flow field measurements and thermal hydraulic analyses in cold moderator vessel

International Nuclear Information System (INIS)

Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsuhiko; Hino, Ryutaro

2001-01-01

The Japan Atomic Energy Research Institute is developing a several MW-scale spallation target system under the High-Intensity Accelerator Project. A cold moderator using supercritical hydrogen is one of the key components in the target system, which directly affects the neutronic performance both in intensity and resolution. Since a hydrogen temperature rise in the moderator vessel affects the neutronic performance, it is necessary to suppress the recirculation and stagnant flows which cause hot spots. In order to develop the conceptual design of the moderator structure in progress, the flow field was measured using a PIV (Particle Image Velocimetry) system under water flow conditions using a flat model that simulated a moderator vessel. From these results, the flow field such as recirculation flows, stagnant flows etc. was clarified. The hydraulic analytical results using the standard k-ε model agreed well with experimental results. Thermal-hydraulic analyses in the moderator vessel were carried out under liquid hydrogen conditions. Based on these results, we clarified the possibility of suppressing the local temperature rise within 3 K under 2 MW operating condition. (author)

Analysis of cracked pressure vessel nozzles by finite elements

International Nuclear Information System (INIS)

Reynen, J.

1975-01-01

In order to assess the safety of pressure vessel nozzles, the analysis should take into account cracks. The paper describes various algorithms, their computer implementations and relative merits to define in an effective way strain energy release rates along the tip front of arbitrary 3 D cracks under arbitary load including thermal strains. These techniques are basically equivalent to substructuring techniques and consequently they can be implemented to only FEM program able to deal with the data handling problems of the substructuring technique. Examples are given carried out with a substructure version of the BERSAFE system. These examples include a corner crack in a pressure vessel nozzle loaded by internal pressure and by thermal stresses. (Auth.)

Simulation of VDE under intervention of vertical stability control and vertical electromagnetic force on the ITER vacuum vessel

International Nuclear Information System (INIS)

Miyamoto, S.; Sugihara, M.; Shinya, K.; Nakamura, Y.; Toshimitsu, S.; Lukash, V.E.; Khayrutdinov, R.R.; Sugie, T.; Kusama, Y.; Yoshino, R.

2012-01-01

Highlights: ► Taking account of intervention of VS control, VDE simulations were carried out. ► Malfunctioning of VS circuit (positive feedback) enhances the vertical force. ► The worst case was explored for vertical force on the ITER vacuum vessel. ► We confirmed the force is still within the design margin even if the worst case. - Abstract: Vertical displacement events (VDEs) and disruptions usually take place under intervention of vertical stability (VS) control and the vertical electromagnetic force induced on vacuum vessels is potentially influenced. This paper presents assessment of the force that arises from the VS control in ITER VDEs using a numerical simulation code DINA. The focus is on a possible malfunctioning of the ex-vessel VS control circuit: radial magnetic field is unintentionally applied to the direction of enhancing the vertical displacement further. Since this type of failure usually causes the largest forces (or halo currents) observed in the present experiments, this situation must be properly accommodated in the design of the ITER vacuum vessel. DINA analysis shows that although the ex-vessel VS control modifies radial field, it does not affect plasma motion and current quench behavior including halo current generation because the vacuum vessel shields the field created by the ex-vessel coils. Nevertheless, the VS control modifies the force on the vessel by directly acting on the eddy current carried by the conducting structures of the vessel. Although the worst case was explored in a range of plasma inductance and pattern of VS control in combination with the in-vessel VS control circuit, the result confirmed that the force is still within the design margin.

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.)

Corrosion of steel tendons used in prestressed concrete pressure vessels

International Nuclear Information System (INIS)

Griess, J.C.; Naus, D.J.

The purpose of this investigation was to determine the corrosion behavior of a high strength steel (ASTM A416-74 grade 270), typical of those used as tensioning tendons in prestressed concrete pressure vessels, in several corrosive environments and to demonstrate the protection afforded by coating the steel with either of two commercial petroleum-base greases or Portland Cement grout. In addition, the few reported incidents of prestressing steel failures in concrete pressure vessels used for containment of nuclear reactors are reviewed. The susceptibility of the steel to stress corrosion cracking and hydrogen embrittlement and its general corrosion rate were determined in several salt solutions. Wires coated with the greases and grout were soaked for long periods in the same solutions and changes in their mechanical properties were subsequently determined. All three coatings appeared to give essentially complete protection but small flaws in the grease coatings were detrimental; flaws or cracks less than 1 mm wide in the grout were without effect

Burnup influence on the VVER-1000 reactor vessel neutron fluence evaluation

International Nuclear Information System (INIS)

Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

2009-01-01

The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of the Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of VVER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in 1/4 depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (authors)

Burnup influence on the WWER-1000 reactor vessel neutron fluence evaluation

International Nuclear Information System (INIS)

Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

2009-01-01

The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of WWER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in ? depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (Authors)

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

R and D Developments. Research Programs on Irradiation Embrittlement of Reactor Vessel Steels

International Nuclear Information System (INIS)

Gomez Briceno, D.; Lapena, J.; Serrano, M.; Perosanz, F.

2000-01-01

Irradiation embrittlement of pressure vessel steels is a degradation mechanism time dependent that can lead to operational restrictions with adverse effects in the efficiency and life of a plant. For the last year, several research programs have been devoted to study thye evaluation of neutronic radiation effect on mechanical properties of pressure vessel steels. However, at the present, there is a growing interest on the development of new methodologies to optimize the surveillance program information, and the understanding of the irradiation damage mechanism. This paper give an overview of international research programs, and on the R+D activities carried out by the Structural Materials Project on irradiation embrittlement on pressure vessel steels. (Author)

Angio-Architectural Features of High-Grade Intracranial Dural Arteriovenous Fistulas: Correlation With Aggressive Clinical Presentation and Hemorrhagic Risk.

Science.gov (United States)

Della Pepa, Giuseppe Maria; Parente, Paolo; D'Argento, Francesco; Pedicelli, Alessandro; Sturiale, Carmelo Lucio; Sabatino, Giovanni; Albanese, Alessio; Puca, Alfredo; Fernandez, Eduardo; Olivi, Alessando; Marchese, Enrico

2017-08-01

High-grade dural arteriovenous fistulas (dAVFs) can present shunts with very different angio-architectural characteristics. Specific hemodynamic factors may affect clinical history and determine very different clinical courses. To evaluate the relationship between some venous angio-architectural features in high-grade dAVFs and clinical presentation. Specific indicators of moderate or severe venous hypertension were analyzed, such as altered configurations of the dural sinuses (by a single or a dual thrombosis), or overload of cortical vessels (restrictions of outflow, pseudophlebitic cortical vessels, and venous aneurysms). The institutional series was retrospectively reviewed (49 cases), and the pattern of venous drainage was analyzed in relationship with clinical presentation (benign/aggressive/hemorrhage). Thirty-five of 49 cases displayed cortical reflux (high-grade dAVFs). This subgroup displayed a benign presentation in 31.42% of cases, an aggressive in 31.42%, and hemorrhage in 37.14%. Our data confirm that within high-grade dAVFs, 2 distinct subpopulations exist according to severity of clinical presentation. Some indicators we examined showed correlation with aggressive nonhemorrhagic manifestations (outflow restriction and pseudophlebitic cortical vessels), while other showed a correlation with hemorrhage (dual thrombosis and venous aneurysms). Current classifications appear insufficient to identify a wide range of conditions that ultimately determine the organization of the cortical venous drainage. Intermediate degrees of venous congestion correlate better with the clinical risk than the simple definition of cortical reflux. The angiographic aspects of venous drainage presented in this study may prove useful to assess dAVF hemodynamic characteristics and identify conditions at higher clinical risk. Copyright © 2017 by the Congress of Neurological Surgeons

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)

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)

Capacity of Prestressed Concrete Containment Vessels with Prestressing Loss

International Nuclear Information System (INIS)

SMITH, JEFFREY A.

2001-01-01

Reduced prestressing and degradation of prestressing tendons in concrete containment vessels were investigated using finite element analysis of a typical prestressed containment vessel. The containment was analyzed during a loss of coolant accident (LOCA) with varying levels of prestress loss and with reduced tendon area. It was found that when selected hoop prestressing tendons were completely removed (as if broken) or when the area of selected hoop tendons was reduced, there was a significant impact on the ultimate capacity of the containment vessel. However, when selected hoop prestressing tendons remained, but with complete loss of prestressing, the predicted ultimate capacity was not significantly affected for this specific loss of coolant accident. Concrete cracking occurred at much lower levels for all cases. For cases where selected vertical tendons were analyzed with reduced prestressing or degradation of the tendons, there also was not a significant impact on the ultimate load carrying capacity for the specific accident analyzed. For other loading scenarios (such as seismic loading) the loss of hoop prestressing with the tendons remaining could be more significant on the ultimate capacity of the containment vessel than found for the accident analyzed. A combination of loss of prestressing and degradation of the vertical tendons could also be more critical during other loading scenarios

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-12-15

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

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%.

Estimation of embrittlement damage risk at neutron embrittled vessel constructions

International Nuclear Information System (INIS)

Staevski, K.; Madzharov, D.; Detistov, P.; Petrova, T.

1998-01-01

In this work a methodology based on Damage mechanics criteria is proposed. This methodology serves for probability assessment of the brittle damage risk for the neutron embrittled vessel elements. The developed methodology is realised in RISK code and has been verified on the base of tough reliability of the pressure vessel, 'Kozloduy' NPP Unit 2. This investigation has been carried out at the given parameters of the possible defects on the vessel's weld 4 taking into account requirements of the western and Russian standards. The obtained values for ductile to brittle transition temperatures, defining the equipment life-time in the presence of maximal defect, are in good consistence with the experimentally determined ones. The analyses of results show that the pressure vessel of 'Kozloduy' NPP Unit 2 has got a high level of reliability from brittle damage risk point of view and that the western standards give more conservative evaluation. On the bases of the results a conclusion is made that the developed methodology enables analysing the influence of possible defects in the neutron embrittled elements on their to reliability and their remained life-time

High-grade and low-grade gliomas: differentiation by using perfusion MR imaging

International Nuclear Information System (INIS)

Hakyemez, B.; Erdogan, C.; Ercan, I.; Ergin, N.; Uysal, S.; Atahan, S.

2005-01-01

AIM: Relative cerebral blood volume (rCBV) is a commonly used perfusion magnetic resonance imaging (MRI) technique for the evaluation of tumour grade. Relative cerebral blood flow (rCBF) has been less studied. The goal of our study was to determine the usefulness of these parameters in evaluating the histopathological grade of the cerebral gliomas. METHODS: This study involved 33 patients (22 high-grade and 11 low-grade glioma cases). MRI was performed for all tumours by using a first-passage gadopentetate dimeglumine T2*-weighted gradient-echo single-shot echo-planar sequence followed by conventional MRI. The rCBV and rCBF were calculated by deconvolution of an arterial input function. The rCBV and rCBF ratios of the lesions were obtained by dividing the values obtained from the normal white matter of the contralateral hemisphere. For statistical analysis Mann-Whitney testing was carried out. A p value of less than 0.05 indicated a statistically significant difference. Receiver operating characteristic curve (ROC) analysis was performed to assess the relationship between the rCBV and rCBF ratios and grade of gliomas. Their cut-off value permitting discrimination was calculated. The correlation between rCBV and CBF ratios and glioma grade was assessed using Pearson correlation analysis. RESULTS: In high-grade gliomas, rCBV and rCBF ratios were measured as 6.50±4.29 and 3.32±1.87 (mean±SD), respectively. In low-grade gliomas, rCBV and rCBF ratios were 1.69±0.51 and 1.16±0.38, respectively. The rCBV and rCBF ratios for high-grade gliomas were statistically different from those of low-grade gliomas (p 0.05). The cut-off value was taken as 1.98 in the rCBV ratio and 1.25 in the rCBF ratio. There was a strong correlation between the rCBV and CBF ratios (Pearson correlation = 0.830, p<0.05). CONCLUSION: Perfusion MRI is useful in the preoperative assessment of the histopathologicalal grade of gliomas; the rCBF ratio in addition to the rCBV ratio can be incorporated

Studies of fragileness in steels of vessels of BWR reactors

International Nuclear Information System (INIS)

Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

2003-01-01

The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA MARK lll reactor and separately with Ni +3 ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A 2 . (Author)

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)

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

Potential-Flow Predictions of a Semi-Displacement Vessel Including Applications to Calm-Water Broaching

OpenAIRE

Ommani, Babak

2013-01-01

As a marine vehicle’s operational speed increases, hydrodynamic pressure plays an increasingly significant role in carrying the vessel’s weight. The shift of importance from hydrostatic to hydrodynamic pressure may cause a vessel, which is stable at rest or low speeds, to become dynamically unstable at high speeds. The nature of the dynamic instability depends mainly on the vessel’s type and speed. Among high speed vessels, semi-displacement mono-hulls are particularly susceptible to a nonosc...

Further fields of application for prestressed cast iron pressure vessels (PCIV)

International Nuclear Information System (INIS)

Guelicher, L.; Schilling, F.E.

1977-01-01

The redundancy of the prestressing system of prestressed structures as well as the clear separation of sealing and load-carrying functions of prestressed cast iron pressure vessels offer substantial advantages over conventional welded steel pressure vessels. Because of the temperature resistance of cast iron up to 400 0 C it is possible to build prestressed pressure vessels commercially as hot-working structures. The compressive strength of cast iron, which is 25 times as high as that of concrete allows for a very compact design of the PCIV. Further specific properties of the PCIV like pre-fabrication of the vessel in the production plant - made possible by a structure assembled from segments - short assembly periods at the construction site etc., may open more fields of application. - PCIV as pressurized storage tanks for the emergency shut down system in nuclear power stations. - PCIV as high pressure vessel for the chemical industry. - PCIV as energy storage. - PCIV for light water reactors. - PCIV as burst protection. It is concluded that the application of prestressed cast iron promises to be successful where either structures with large volumes and high pressures and/or temperatures are required or where aspects of safety allow for efficient use of prestressed structures. (Auth.)

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.

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.

Modeling Erlang's Ideal Grading with Multirate BPP Traffic

Directory of Open Access Journals (Sweden)

Mariusz Glabowski

2012-01-01

Full Text Available This paper presents a complete methodology for modeling gradings (also called non-full-availability groups servicing single-service and multi-service traffic streams. The methodology worked out by the authors makes it possible to determine traffic characteristics of various types of gradings with state-dependent call arrival processes, including a new proposed structure of the Erlang’s Ideal Grading with the multirate links. The elaborated models of the gradings can be used for modeling different systems of modern networks, for example, the radio interfaces of the UMTS system, switching networks carrying a mixture of different multirate traffic streams, and video-on-demand systems. The results of the analytical calculations are compared with the results of the simulation data for selected gradings, which confirm high accuracy of the proposed methodology.

Effect of non-Newtonian characteristics of blood on magnetic particle capture in occluded blood vessel

Science.gov (United States)

Bose, Sayan; Banerjee, Moloy

2015-01-01

Magnetic nanoparticles drug carriers continue to attract considerable interest for drug targeting in the treatment of cancer and other pathological conditions. Magnetic carrier particles with surface-bound drug molecules are injected into the vascular system upstream from the desired target site, and are captured at the target site via a local applied magnetic field. Herein, a numerical investigation of steady magnetic drug targeting (MDT) using functionalized magnetic micro-spheres in partly occluded blood vessel having a 90° bent is presented considering the effects of non-Newtonian characteristics of blood. An Eulerian-Lagrangian technique is adopted to resolve the hemodynamic flow and the motion of the magnetic particles in the flow using ANSYS FLUENT. An implantable infinitely long cylindrical current carrying conductor is used to create the requisite magnetic field. Targeted transport of the magnetic particles in a partly occluded vessel differs distinctly from the same in a regular unblocked vessel. Parametric investigation is conducted and the influence of the insert configuration and its position from the central plane of the artery (zoffset), particle size (dp) and its magnetic property (χ) and the magnitude of current (I) on the "capture efficiency" (CE) is reported. Analysis shows that there exists an optimum regime of operating parameters for which deposition of the drug carrying magnetic particles in a target zone on the partly occluded vessel wall can be maximized. The results provide useful design bases for in vitro set up for the investigation of MDT in stenosed blood vessels.

Fracture toughness of irradiated and recovered vessel steels

International Nuclear Information System (INIS)

Perosanz, F.; Lapena, J.

1998-01-01

This paper presents the fracture toughness measurements carried out on three vessel steels in an irradiated condition and after a post-irradiation recovery treatment. A statistical approach and the fracture parameters corresponding to two theoretical models of the fracture tests are used for evaluating toughness. Test results show that the neutron fluence gradually transforms the fracture behaviour of the vessel steels from ductile to brittle and seriously reduces their fracture toughness. The effectiveness of the recovery treatment, as evaluated from the toughness measurements, is confirmed, although the efficiency is not the same for the steels and depends on the evaluation parameter except in the case of almost complete recovery. The recovery effect increases with the received neutron fluence if the toughness values after treatment are compared with those in the irradiated condition rather than those in the as received condition. (orig.)

A Review of Algorithms for Retinal Vessel Segmentation

Directory of Open Access Journals (Sweden)

Monserrate Intriago Pazmiño

2014-10-01

Full Text Available This paper presents a review of algorithms for extracting blood vessels network from retinal images. Since retina is a complex and delicate ocular structure, a huge effort in computer vision is devoted to study blood vessels network for helping the diagnosis of pathologies like diabetic retinopathy, hypertension retinopathy, retinopathy of prematurity or glaucoma. To carry out this process many works for normal and abnormal images have been proposed recently. These methods include combinations of algorithms like Gaussian and Gabor filters, histogram equalization, clustering, binarization, motion contrast, matched filters, combined corner/edge detectors, multi-scale line operators, neural networks, ants, genetic algorithms, morphological operators. To apply these algorithms pre-processing tasks are needed. Most of these algorithms have been tested on publicly retinal databases. We have include a table summarizing algorithms and results of their assessment.

Gelling agents and culture vessels affect in vitro multiplication of banana plantlets.

Science.gov (United States)

Kaçar, Y A; Biçen, B; Varol, I; Mendi, Y Y; Serçe, S; Cetiner, S

2010-03-09

Agar is the most commonly used gelling agent in media for plant tissue culture. Because of the high price of tissue-culture-grade agar, attempts have been made to identify suitable alternatives. The type of culture vessel and lid also affects the gaseous composition inside the vessel as well as light penetration. In turn, the vessel affects growth parameters, such as shoot elongation, proliferation and fresh weight, as well as hyperhydric degradation processes. We examined the effects of different culture vessels, including commercial glass jars, magenta boxes, and disposable containers, as well as different gelling agents (agar-agar, Agargel, Phytagel, and plant agar) on the micropropagation of Dwarf Cavendish bananas in an effort to find a combination that yields large numbers of high-quality seedlings. The different culture vessels did not significantly affect seedling culture success. The medium significantly affected shoot weight. Phytagel resulted in the highest shoot weight (overall mean = 2.4 g), while agar, Agargel and plant agar resulted in 1.7, 2.2 and 2.2 g, respectively. Disposable container/Phytagel and Magenta/Agargel combinations yielded the highest shoot weights (2.9 and 3.0 g, respectively). Mean shoot length increased progressively with subculture (four subcultures were made). The highest mean shoot length was obtained with Phytagel and Agargel media (6.4 and 6.3 cm, respectively). Shoot number was significantly affected by medium only at subculture 4. Overall, the highest mean shoot length was obtained with the Magenta/Agargel combination (8.5 cm). Phytagel and plant agar gave higher mean shoot number than agar and Agargel (2.1, 2.1 and 1.7 and 1.9, respectively). The costs of the media and of the culture vessels need to be taken into account for final choice of the banana shoot culture system.

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

Review of the Conceptual Design for In-Vessel Fuel Handling Machines in SFR

International Nuclear Information System (INIS)

Kim, S. H.; Koo, G. H.

2012-01-01

The main in-vessel fuel handling machines in sodium cooled fast reactor(SFR) are composed of the in-vessel transfer machine(IVTM) and the rotating plug. These machines perform the function to handle fuel assemblies inside the reactor core during the refueling time. The IVTM should be able to access all areas above the reactor core and the fuel transfer port which can discharge the fuel assembly by the rotation of the rotating plug. In the 600 MWe demonstration reactor, the conceptual design of the in-vessel fuel handling machines was carried out. As shown in Fig. 1, the invessel fuel handling machines of the demonstration reactor are the double rotating plug type. With reference to the given core configuration of the demonstration reactor, the arrangement design of the rotating plug was carried out by using the developed simulation program. At present, the conceptual design of SFR prototype reactor which has small capacity of about 100 MWe is being started. Thus, it is necessary the economical efficiency and the reliability of the in-vessel fuel handling machines are reviewed according to the reduction of the power capacity. In this study, the preliminary design concepts of the main invessel fuel handling machines according to the fuel handling type are compared. Also, the design characteristics for the driving mechanism of the IVTM in the demonstration reactor and the recovery concept from the malfunction are reviewed

Design, fabrication and quality assurance of pressure vessels

International Nuclear Information System (INIS)

Kimura, Ichiro; Miki, Masao; Yamazaki, Tsuneji; Tanaka, Yoshikazu; Sato, Misao

1978-01-01

The production facilities, design and manufacturing technologies, and quality assurance in the Toyo Works, Ehime Manufactory, Sumitomo Heavy Industries, Ltd., which manufactures pressure vessels, are described, and especially the actual example of non-destructive tests is shown. The Toyo Works was completed in April, 1973, to manufacture large structures such as pressure vessels, offshore structures and bridges. The total area of the site is 535,000 m 2 , that of factory buildings is 33,600 m 2 , and the outdoor assembling yard is 114,800 m 2 . The large dry dock and main installations such as 12,000 tf hydraulic press, an annealing furnace, a heat treating furnace, a quenching tank, a horizontal boring machine, 6 m vertical lathe, various welding machines, 8 MeV X-ray apparatus, sand blasting and pickling facilities, and two 160 t cranes for shipment are arranged so as to enable smooth flow of production. The standards for chemical pressure vessels in various countries are compared, and considerably high allowable stress is adopted in Europe. The design and stress analysis of pressure vessels are carried out in accordance with ASME Section 8, Div. 1 or Div. 2. As for the materials, attention must be paid to the change of properties due to heat and strain, temper brittleness, low temperature toughness and so on. The quality assurance system must be established to observe the requirements of standards. (Kako, I.)

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

Adolescents Carrying Handguns and Taking Them to School: Psychosocial Correlates among Public School Students in Illinois.

Science.gov (United States)

Williams, Sunyna S.; Mulhall, Peter F.; Reis, Janet S.; DeVille, John O.

2002-01-01

Examines psychosocial correlates of adolescents carrying a handgun and taking a handgun to school. Survey participants were approximately 22,000 6th, 8th, and 10th grade public school students from Illinois. Results showed that the strongest correlates of handgun carrying behaviors were variables directly associated with handguns and violence,…

Conceptual design of the handling and storage system of the spent target vessel for neutron scattering facility 2

International Nuclear Information System (INIS)

Adachi, Junichi; Kaminaga, Masanori; Sasaki, Shinobu; Haga, Katsuhiro; Aso, Tomokazu; Kinoshita, Hidetaka; Hino, Ryutaro

2002-01-01

In designing the neutron scattering facility, a spent target vessel should be replaced with remote handling devices in order to protect radioactive exposure, since it would be highly activated through the high energy neutron irradiation caused by the spallation reaction between mercury of the target material and the MW-class proton beam. In the storage of the spent target vessel, it is necessary to consider decay heat of the target vessel and mercury contamination caused by vaporization of the residual mercury in the vessel. A conceptual design has been carried out to establish basic concept and to clarify its specification of main equipments on handling and storage systems for the spent target vessel. This report presents the basic concept and a system plot plan based on latest design works of remote handling devices such as a spent target vessel storage cask and a target vessel exchange trolley, which aim at reasonability and simplification. In addition, storage systems for the spent moderator vessel, the spent proton beam window and the spent reflector vessel are also investigated based on the plot plan. (author)

Evaluation of thermal ratcheting of reactor vessel wall near the sodium surface

International Nuclear Information System (INIS)

Take, Kohji; Fujioka, Terutaka; Yano, Kazutaka

1989-01-01

Plastic ratcheting of reactor vessels may occur by an axially moving thermal gradient without primary stress. So there is a need to establish a proper prediction method for the plastic ratcheting. In this study, inelastic FEM analyses of reactor vessel model by using an advanced constitutive equation were carried out in order to comprehend plastic ratcheting behaviour of cylinder which subject to an axially moving thermal gradient. As a result of analyses, a basic mechanism of this ratcheting was found. And it also indicated that cyclic hardening behaviour will became important for development of evaluation method. (author)

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

Evaluation of carotid vessel wall enhancement with image subtraction after gadobenate dimeglumine-enhanced MR angiography

International Nuclear Information System (INIS)

Sardanelli, Francesco; Di Leo, Giovanni; Aliprandi, Alberto; Flor, Nicola; Papini, Giacomo D.E.; Roccatagliata, Luca; Cotticelli, Biagio; Nano, Giovanni; Cornalba, Gianpaolo

2009-01-01

Objectives: This study was aimed at testing the value of image subtraction for evaluating carotid vessel wall enhancement in contrast-enhanced MR angiography (MRA). Materials and methods: IRB approval was obtained. The scans of 81 consecutive patients who underwent carotid MRA with 0.1 mmol/kg of gadobenate dimeglumine were reviewed. Axial carotid 3D T1-weighted fast low-angle shot sequence before and 3 min after contrast injection were acquired and subtracted (enhanced minus unenhanced). Vessel wall enhancement was assigned a four-point score using native or subtracted images from 0 (no enhancement) to 3 (strong enhancement). Stenosis degree was graded according to NASCET. Results: With native images, vessel wall enhancement was detected in 20/81 patients (25%) and in 20/161 carotids (12%), and scored 2.0 ± 0.6 (mean ± standard deviation); with subtracted images, in 21/81 (26%) and 22/161 (14%), and scored 2.5 ± 0.6, respectively (P < 0.001, Sign test). The overall stenosis degree distribution was: mild, 41/161 (25%); moderate, 77/161 (48%); severe, 43/161 (27%). Carotids with moderate stenosis showed vessel wall enhancement with a frequency (17/77, 22%) significantly higher than that observed in carotids with mild stenosis (1/41, 2%) (P = 0.005, Fisher exact test) and higher, even though with borderline significance (P = 0.078, Fisher exact test), than that observed in carotids with severe stenosis (4/43, 9%). Conclusion: Roughly a quarter of patients undergoing carotid MRA showed vessel wall enhancement. Image subtraction improved vessel wall enhancement conspicuity. Vessel wall enhancement seems to be an event relatively independent from the degree of stenosis. Further studies are warranted to define the relation between vessel wall enhancement and histopathology, inflammatory status, and instability.

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)

Utility of Vascular Enhancement Technique (ClarifyTM) in Ultrasonographic Evaluation of Abdominal Vessels

International Nuclear Information System (INIS)

Oh, Jong Young; Cho, Jin Han; Choi, Jong Cheol; Shin, Tae Beom; Lee, Jin Hwa; Yoon, Seong Kuk; Nam, Kyung Jin

2006-01-01

Vascular enhancement (VE) technology(ClarifyTM) is a new technique in vascular, B-mode imaging. The purpose of this study was to evaluate the value of VE technology in ultrasonographic diagnosis of abdominal vasculature. Seventy-one adult patients (39 men and 32 women: age range, 25-89 years: mean age, 56 years) who had undergone abdominal ultrasonography were included in this study. The imaging was performed with a 1.8-4.0 MHz convex array transducer (SONOLINE, Antares, Siemens Medical Solutions, WA) by an abdominal radiologist. The radiologist obtained images of the same vascular area with each of conventional ultrasonography imaging (CUS), tissue harmonic imaging (THI), CUS plus VE technique and THI plus VE technique. Images were divided into normal (56) and abnormal (15) groups. The vessel visibility, conspicuity of the vascular wall and contrast resolution with adjacent structures were evaluated in the normal group, and the lesion conspicuity and border sharpness were evaluated in the abnormal group. On the PACS monitor, the images were graded into four grades by two radiologists in consensus. Statistical analysis was performed using Wilcoxon signed rank test. In the normal group, all parameters of the ultrasonographic imaging which applied the VE technique were superior to those of the imaging without VE technique (p < 0.05). In the abnormal group, combined use of VE technique with CUS or THI provided better results than CUS or THI alone in terms of lesion conspicuity and border sharpness (p < 0.05). THI combined with VE technique provided the best image quality among the 4 ultrasonographic methods examined in this study for the evaluation of both normal and abnormal abdominal vessels (p < 0.05). VE technology was a helpful technique to evaluate the abdominal vasculature. Furthermore, VE technique combined with THI provided better image quality than other ultrasonographic methods in the evaluation of abdominal vessels

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

SURFACE TREATMENT AND EXAMINATION OF GRADE 2 AND GRADE 5 TITANIUM

Directory of Open Access Journals (Sweden)

Peter Nagy

2016-02-01

Full Text Available Surface characteristics play an important role in the implant-bone integration that is required for the long-term reliability of dental and orthopedic implants. In this paper, we investigate the effect of acid etching on the mass reduction and roughness of grade 2 and grade 5 Ti under controlled experimental conditions. Three different etching compounds were investigated: 30% HCl, 85% H3PO4 and the compound of 30% (COOH2 × 2H2O and 30% H2O2 in various treatment intervals under controlled temperature. Stereo microscopy, scanning electron microscopy, roughness and weight measurements were carried out on the samples. We found that neither 85% H3PO4 nor the compound of 30% (COOH2 × 2H2O and 30% H2O2 were able to remove the machining marks from the surface of Ti discs in our experimental setting. On the other hand, etching in 30% HCl yielded even surfaces both on Ti grade 2 and 5 discs. We also found that etching at higher temperatures in 30% HCl resulted in significant mass loss.

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...

Design and issues of the ITER in-vessel components: ITER Joint central team and home teams

International Nuclear Information System (INIS)

Parker, R.R.

1998-01-01

This paper surveys the status of the design of the in-vessel components for ITER, in particular the major components, namely the vacuum vessel, blanket and first wall, and divertor, and the interface of selected ancillary systems such as those used for RF heating and current drive, and for diagnostics. The vacuum vessel is a double-walled structure constructed from two toroidal shells joined by ribs. The space between the skins is filled with shield plates directly cooled by water. The structural material is 316 LN IG (ITER grade). Toroidal supports joining the vessel midplane ports with the TF structure limit possible differential toroidal displacements, as might occur due to seismic or vertical displacement events (VDEs). A variety of load conditions corresponding to normal and off-normal loads have been considered and in all cases peak vessel stresses are within allowables. The blanket system consists of approximately 700 modules, each weighing ∝4 t. The integrated first wall consists of a beryllium-tiled copper mat bonded to the water-cooled SS shield block. The copper mat functions as a heat sink and has imbedded in it an array of SS tubes providing water cooling. The modules are mechanically attached to a toroidal backplate. Loads due to centered disruptions are reacted via hoop stress in the backplate, whereas net vertical and horizontal loads such as those arising from VDEs are transferred through the backplate and divertor supports to the vessel. (orig.)

Cast iron as structural material for hot-working reactor vessels (PCIV)

International Nuclear Information System (INIS)

Ostendorf, H.; Schmidt, G.; Pittack, W.

1977-01-01

Cast iron with lamellar graphite is best suited for prestressed structures, because its compressive strength is nearly 4 times its tensile strength. In comparison to room temperature, cast iron with lamellar graphite shows essentially no loss of strength up to temperatures of 400 0 C. Under the particular aspect to use cast iron for hot-working prestressed reactor pressure vessels (PCIV) (Prestressed cast iron vessel=PCIV) a materials testing program is carried out, which meets the strict certification requirements for materials in the construction of reactor pressure vessels and which completes the presently available knowledge of cast iron. Especially in the following fields an extension and supplement of the present level of knowledge is necessary. - Mechanical properties under compressive stresses. - Material properties at elevated temperatures. - Influence of irradiation on mechanical and physical properties. - Production standards and quality control. The state of the research and the available data of the material testing program are reported. (Auth.)

Cast iron as structural material for hot-working reactor vessels (PCIV)

International Nuclear Information System (INIS)

Ostendorf, H.; Schmidt, G.; Pittack, W.

1977-01-01

Cast iron with lamellar graphite is best suited for prestressed structures, because its compressive strength is nearly 4 times its tensile strength. In comparison to room temperature, cast iron with lamellar graphite shows essentially no loss of strength up to temperatures of 400 0 C. Under the particular aspect to use cast iron for hot-working prestressed reactor pressure vessels (PCIV) (Prestressed cast iron vessel=PCIV) a materials testing program is carried out, which meets the strict certification requirements for materials in the construction of reactor pressure vessels and which completes the presently available knowledge of cast iron. Especially in the following fields an extension and supplement of the present level of knowledge is necessary: mechanical properties under compressive stresses; material properties at elevated temperatures; influence of irradiation on mechanical and physical properties; production standards and quality control. The state of the research and the available data of the material testing program are reported

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

Improvement to reactor vessel

International Nuclear Information System (INIS)

1974-01-01

The vessel described includes a prestressed concrete vessel containing a chamber and a removable cover closing this chamber. The cover is in concrete and is kept in its closed position by main and auxiliary retainers, comprising fittings integral with the concrete of the vessel. The auxiliary retainers pass through the concrete of the cover. This improvement may be applied to BWR, PWR and LMFBR type reactor vessel [fr

Investigation of vessel traffic passing through the Aleutian Islands with discussion of risk factors that could lead to oil pollution

International Nuclear Information System (INIS)

Eley, D.

2006-01-01

The oil spill risks posed by marine vessels travelling the North Pacific route from North America to Asia were discussed with reference to the grounding and break-up of the bulk grain ship M/V Selendang Ayu at Unalaska Island in Alaska's Aleutian Islands. The challenge of gathering and categorizing vessel traffic data in the Aleutians was also discussed along with a review of methods for developing an accurate traffic study for this large, remote maritime region. The travelling route passes through, or lies in close proximity to large and valuable commercial fishing grounds as well as the Alaska Maritime National Wildlife Refuge. As most vessels remain in international waters, they do not report their presence to state and national authorities and are exempt from contingency planning requirements. This paper listed the factors affecting risk of damage from oil spills from marine vessels. These include the volume of oil carried; types of oil; proximity to environmentally sensitive areas; fate of spill; location of spill response equipment; number of vessels travelling through the area; time that vessels are in the area; type and age of vessels; environmental factors affecting sailing conditions; factors limiting rescue; and interaction with regulatory agencies. When considered as a whole, these factors can help in deciding the degree and type of contingency planning. In order to estimate traffic through the North Pacific route, the authors compared, combined and extrapolated information from similar datasets. The task involved 3 steps: (1) estimating the number of trans-Pacific voyages through the Aleutians, (2) estimating vessel type, and (3) estimating fuel oil carried by vessel type. It was determined that more than 2,700 ship voyages pass through the Aleutians every year, of which 50 carry a total of 800 million gallons of oil as cargo. It was noted that serious ship accidents occur so infrequently in remote areas that it was impossible to establish an accurate

Low-grade myofibroblastic sarcoma arising in fibroadenoma of the breast-A case report.

Science.gov (United States)

Myong, Na-Hye; Min, Jun-Won

2016-03-25

Myofibroblastic sarcoma or myofibrosarcoma is a malignant tumor of myofibroblasts and known to develop rarely in the breast, but its underlying lesion and tumor cell origin have never been reported yet. A 61-year-old female presented with a gradually growing breast mass with well-demarcated ovoid nodular shape. The tumor was histologically characterized by fascicular-growing spindle cell proliferation with large areas of hyalinized fibrosis and focally ductal epithelial remnants embedded in myxoid stroma, mimicking a fibroadenomatous lesion. It had frequent mitoses of 5-16/10 high-power fields, hemorrhagic necrosis, and focally pericapsular invasion. The spindle cells were diffusely immunoreactive for fibronectin, smooth muscle actin, and calponin, which suggest a myofibroblastic origin. Multiple irregularly thickened vessels with medial or pericytic cell proliferation were found to be merged with the intrinsic tumor cells. The tumor could be diagnosed low-grade myofibroblastic sarcoma arising in an old fibroadenoma. We report a case of a low-grade mammary myofibrosarcoma that showed a background lesion of fibroadenoma first in the worldwide literature and suggest the pericytes or medial muscle cells of the intratumoral vessels as the cell origin of the myofibroblastic sarcoma.

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

In-vessel tritium

International Nuclear Information System (INIS)

Ueda, Yoshio; Ohya, Kaoru; Ashikawa, Naoko; Ito, Atsushi M.; Kato, Daiji; Kawamura, Gakushi; Takayama, Arimichi; Tomita, Yukihiro; Nakamura, Hiroaki; Ono, Tadayoshi; Kawashima, Hisato; Shimizu, Katsuhiro; Takizuka, Tomonori; Nakano, Tomohide; Nakamura, Makoto; Hoshino, Kazuo; Kenmotsu, Takahiro; Wada, Motoi; Saito, Seiki; Takagi, Ikuji; Tanaka, Yasunori; Tanabe, Tetsuo; Yoshida, Masafumi; Toma, Mitsunori; Hatayama, Akiyoshi; Homma, Yuki; Tolstikhina, Inga Yu.

2012-01-01

The in-vessel tritium research is closely related to the plasma-materials interaction. It deals with the edge-plasma-wall interaction, the wall erosion, transport and re-deposition of neutral particles and the effect of neutral particles on the fuel recycling. Since the in-vessel tritium shows a complex nonlinear behavior, there remain many unsolved problems. So far, behaviors of in-vessel tritium have been investigated by two groups A01 and A02. The A01 group performed experiments on accumulation and recovery of tritium in thermonuclear fusion reactors and the A02 group studied theory and simulation on the in-vessel tritium behavior. In the present article, outcomes of the research are reviewed. (author)

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

International Nuclear Information System (INIS)

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

1990-01-01

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

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

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)

Hand-eye coordinative remote maintenance in a tokamak vessel

Energy Technology Data Exchange (ETDEWEB)

Qiu, Qiang, E-mail: qiu6401@sjtu.edu.cn; Gu, Kai, E-mail: gukai0707@sjtu.edu.cn; Wang, Pengfei, E-mail: wpf790714@163.com; Bai, Weibang, E-mail: 654253204@qq.com; Cao, Qixin, E-mail: qxcao@sjtu.edu.cn

2016-03-15

Highlights: • If there is not rotation between the visual coordinate frame (O{sub e}X{sub e}Y{sub e}) and hand coordinate frame (O{sub h}X{sub h}Y{sub h}), a person can coordinate the movement between hand and eye easily. • We establish an alignment between the movement of the operator's hand and the visual scene of the end-effector as displayed on the monitor. • A potential function is set up in a simplified vacuum vessel model to provide a fast collision checking, and the alignment between repulsive force and Omega 7 feedback force is accomplished. • We carry out an experiment to evaluate its performance in a remote handling task. - Abstract: The reliability is vitally important for the remote maintenance in a tokamak vessel. In order to establish a more accurate and safer remote handling system, a hand-eye coordination method and an artificial potential function based collision avoidance method were proposed in this paper. At the end of this paper, these methods were implemented to a bolts tightening maintenance task, which was carried out in our 1/10 scale tokamak model. Experiment results have verified the value of the hand-eye coordination method and the collision avoidance method.

Hand-eye coordinative remote maintenance in a tokamak vessel

International Nuclear Information System (INIS)

Qiu, Qiang; Gu, Kai; Wang, Pengfei; Bai, Weibang; Cao, Qixin

2016-01-01

Highlights: • If there is not rotation between the visual coordinate frame (O_eX_eY_e) and hand coordinate frame (O_hX_hY_h), a person can coordinate the movement between hand and eye easily. • We establish an alignment between the movement of the operator's hand and the visual scene of the end-effector as displayed on the monitor. • A potential function is set up in a simplified vacuum vessel model to provide a fast collision checking, and the alignment between repulsive force and Omega 7 feedback force is accomplished. • We carry out an experiment to evaluate its performance in a remote handling task. - Abstract: The reliability is vitally important for the remote maintenance in a tokamak vessel. In order to establish a more accurate and safer remote handling system, a hand-eye coordination method and an artificial potential function based collision avoidance method were proposed in this paper. At the end of this paper, these methods were implemented to a bolts tightening maintenance task, which was carried out in our 1/10 scale tokamak model. Experiment results have verified the value of the hand-eye coordination method and the collision avoidance method.

On the Adequacy of API 521 Relief-Valve Sizing Method for Gas-Filled Pressure Vessels Exposed to Fire

Directory of Open Access Journals (Sweden)

Anders Andreasen

2018-03-01

Full Text Available In this paper, the adequacy of the legacy API 521 guidance on pressure relief valve (PRV sizing for gas-filled vessels subjected to external fire is investigated. Multiple studies show that in many cases, the installation of a PRV offers little or no protection—therefore provides an unfounded sense of security. Often the vessel wall will be weakened by high temperatures, before the PRV relieving pressure is reached. In this article, a multiparameter study has been performed taking into consideration various vessel sizes, design pressures (implicitly vessel wall thickness, vessel operating pressure, fire type (pool fire or jet fire by applying the methodology presented in the Scandpower guideline. A transient thermomechanical response analysis has been carried out to accurately determine vessel rupture times. It is demonstrated that only vessels with relatively thick walls, as a result of high design pressures, benefit from the presence of a PRV, while for most cases no appreciable increase in the vessel survival time beyond the onset of relief is observed. For most of the cases studied, vessel rupture will occur before the relieving pressure of the PRV is reached.

Manufacture of electron beam irradiation vessel and its characteristics

International Nuclear Information System (INIS)

Kanazawa, Takao; Haruyama, Yasuyuki; Yotsumoto, Keiichi

1992-05-01

Electron beam irradiation vessel, which is used for the irradiation of samples under an inert or a vacuum atmosphere, is made by considering the temperature control during or after irradiation. The vessel was composed of the temperature controlable samples supporting plate, beam slit with water cooling plate and the insert of thermosensor. The four samples supporting plate was produced with the materials made up of aluminium, stainless steel (SUS304), and copper. The stainless steel supporting plate has a heater inside the cooling pipes for the high temperature treatment of samples without exposure to atmosphere after the irradiation. In this report, the temperature distribution and dose characteristics such as dose distribution and effects of backscattered electron were studied by using several supporting plate and the comparison of the experimental results with the simulated results was also carried out. (author)

NCSX Vacuum Vessel Fabrication

International Nuclear Information System (INIS)

Viola ME; Brown T; Heitzenroeder P; Malinowski F; Reiersen W; Sutton L; Goranson P; Nelson B; Cole M; Manuel M; McCorkle D.

2005-01-01

The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in conjunction with the Oak Ridge National Laboratory (ORNL). The goal of this experiment is to develop a device which has the steady state properties of a traditional stellarator along with the high performance characteristics of a tokamak. A key element of this device is its highly shaped Inconel 625 vacuum vessel. This paper describes the manufacturing of the vessel. The vessel is being fabricated by Major Tool and Machine, Inc. (MTM) in three identical 120 o vessel segments, corresponding to the three NCSX field periods, in order to accommodate assembly of the device. The port extensions are welded on, leak checked, cut off within 1-inch of the vessel surface at MTM and then reattached at PPPL, to accommodate assembly of the close-fitting modular coils that surround the vessel. The 120 o vessel segments are formed by welding two 60 o segments together. Each 60 o segment is fabricated by welding ten press-formed panels together over a collapsible welding fixture which is needed to precisely position the panels. The vessel is joined at assembly by welding via custom machined 8-inch (20.3 cm) wide spacer ''spool pieces''. The vessel must have a total leak rate less than 5 X 10 -6 t-l/s, magnetic permeability less than 1.02(micro), and its contours must be within 0.188-inch (4.76 mm). It is scheduled for completion in January 2006

SCF analysis of a pressurized vessel-nozzle intersection with wall thinning damage

International Nuclear Information System (INIS)

Qadir, M.; Redekop, D.

2009-01-01

A three-dimensional finite element analysis is carried out of a pressurized vessel-nozzle intersection (tee joint), with wall thinning damage. A convergence-validation study is first carried out for undamaged intersections, in which comparisons are made with previously published work for the stress concentration factor (SCF), and good agreement is observed. A study is then carried out for specific tee joints to examine the effect on the SCF of varying the extent of the wall thinning damage. Finally, a parametric study is conducted in which the SCF is computed for a wide range of tee joints, initially considered undamaged, and then with wall thinning damage.

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...

The design of lifting attachments for the erection of large diameter and heavy wall pressure vessels

International Nuclear Information System (INIS)

Antalffy, Leslie P.; Miller, George A.; Kirkpatrick, Kenneth D.; Rajguru, Anil; Zhu, Yong

2016-01-01

Lifting attachments for the erection of large diameter and heavy wall pressure vessels require special consideration to ensure that their attachment to their vessel shells or heads do not overstress the vessel during the erection process when lifting these from grade onto their respective foundations. Today, in refinery and petrochemical services, large diameter vessels with diameters ranging up to 15 m and reactors with lifting weights in the range of 700–1400 tons are not uncommon. In today's fabrication market, these vessels may be purchased and fabricated in shops dispersed globally and will require unique equipment for their safe handling, transportation and subsequent erection. The challenge is to design the lifting attachments in such a manner that the attachments provide a safe, cost effective and effective solution based upon the limitations of the job site lift equipment available for erection. Such equipment for the transportation and subsequent lifting of large diameter and heavy wall pressure equipment is usually scarce and quite expensive. Planning ahead, well in advance of the lift date is almost a mandatory requirement. Usually, the specific parameters of the vessel to be lifted and the lifting equipment available at the site will dictate the type of lifting attachments to be designed for the vessel. Once the type of vessel attachment has been chosen, careful consideration must be given to the design of attachments to the pressure vessel in consideration to ensure that the vessel and lifting components are not overstressed during the lifting process. The paper also discusses different types of lifting attachments that may be attached to each end of the vessel either by bolting or welding and discusses the pros and cons of each. The paper also provides an example of a finite element analysis (FEA) of a top nozzle, a FEA of a pair of lifting trunnions and a FEA of welded on lifting lugs for buried pipe. The purpose of the paper is to outline the

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 ...

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.

Evaluation of structural reliability for vacuum vessel under external pressure and electromagnetic force

International Nuclear Information System (INIS)

Minato, Akio

1983-08-01

Static and dynamic structural analyses of the vacuum vessel for a Swimming Pool Type Tokamak Reactor (SPTR) have been conducted under the external pressure (hydraulic and atmospheric pressure) during normal operation or the electromagnetic force due to plasma disruption. The reactor structural design is based on the concept that the adjacent modules of the vacuum vessel are not connected mechanically with bolts in the torus inboard region each other, so as to save the required space for inserting the remote handling machine for tightenning and untightenning bolts in the region and to simplify the repair and maintenance of the reactor. The structural analyses of the vacuum vessel have been carried out under the external pressure and the electromagnetic force and the structural reliability against the static and dynamic loads is estimated. The several configurations of the lip seal between the modules, which is required to make a plasma vacuum boundary, have been proposed and the structural strength under the forced displacements due to the deformation of the vacuum vessel is also estimated. (author)

46 CFR 199.620 - Alternatives for all vessels in a specified service.

Science.gov (United States)

2010-10-01

... that are readily accessible to each watch or work station, the requirement in § 199.70(b)(2)(iv) to have lifejackets at each watch or work station need not be met. (2) If the vessel carries lifejackets... may use lights for lifejackets and immersions suits approved under series 161.012. However, lifejacket...

Concept of a nuclear powered submersible research vessel and a compact reactor

International Nuclear Information System (INIS)

Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Ishida, Toshihisa; Nishimura, Hajime; Tokunaga, Sango

2001-07-01

A conceptual design study of a submersible research vessel navigating in 600 m depth and a compact nuclear reactor were carried out for the expansion of the nuclear power utilization. The mission of the vessel is the research of mechanism of the climate change to predict the global environment. Through conditions of the Arctic Ocean and the sea at high latitude have significant impacts on the global environmental change, it is difficult to investigate those areas by ordinary ships because of thick ice or storm. Therefore the research vessel is mainly utilized in the Arctic Ocean and the sea at high latitude. By taking account of the research mission, the basic specifications of the vessel are decided; the total weight is 500 t, the submersible depth is 600 m, the maximum speed is 12 knots (22.2 km/h), and the number of crews is 16. Nuclear power has an advantage in supplying large power of electricity in the sea for long period. Based on the requirements, it has been decided that two sets of submersible compact reactor, SCR, which is light-weighted and of enhanced safety characteristics of supply the total electricity of 500 kW. (author)

Positioning means for circumferentially locating inspection apparatus in a nuclear reactor vessel

International Nuclear Information System (INIS)

Burns, D.C.

1979-01-01

Positioning means for locating inspection apparatus used to volumetrically examine a nuclear reactor vessel is disclosed. The positioning means is provided with a support ring having an annular key positioned longitudinally about its periphery. Three support legs are attached to the support ring by brackets adapted to fit the annular key. The support ring also carries three guide stud bushings which are movably mounted by clamps adapted to engage the support ring key. Prior to lowering the inspection apparatus into the vessel, the guide stud bushings are each moved to a point of alignment with one of three guide studs extending upwardly from the vessel. After alignment has been verified, the guide stud bushings are clamped in position. The inspection apparatus is lowered towards its fully seated position within the vessel and is coarsely circumferentially positioned with by the engagement of the guide studs within the guide stud bushings. A fine degree of circumferential positioning is achieved by providing a specially configured shoe for one of the support legs. With the core barrel internals in, the special shoe is adapted to key onto a core barrel pin the exact location of which is known. With the core barrel internals removed, the special shoe is adapted to place a locating key into a notch in a vessel flange, the location of which is known. As the inspection apparatus is lowered into its fully seated position, exact circumferential positioning with respect to the vessel is achieved. The other support legs rest on an inner circumferential flange so that no portion of the inspection apparatus touches or threatens the vessel's top flange. 19 claims

Effect of Heat Treatment on Microstructure and Hardness of Grade 91 Steel

Directory of Open Access Journals (Sweden)

Triratna Shrestha

2015-01-01

Full Text Available Grade 91 steel (modified 9Cr-1Mo steel is considered a prospective material for the Next Generation Nuclear Power Plant for application in reactor pressure vessels at temperatures of up to 650 °C. In this study, heat treatment of Grade 91 steel was performed by normalizing and tempering the steel at various temperatures for different periods of time. Optical microscopy, scanning and transmission electron microscopy in conjunction with microhardness profiles and calorimetric plots were used to understand the microstructural evolution including precipitate structures and were correlated with mechanical behavior of the steel. Thermo-Calc™ calculations were used to support the experimental work. Furthermore, carbon isopleth and temperature dependencies of the volume fraction of different precipitates were constructed.

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

Identification and assessment of BWR in-vessel severe accident mitigation strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Kress, T.S.; Cleveland, J.C.; Petek, M.

1992-01-01

This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident management. These results are described in detail in the just-released report Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, NUREG/CR-5869, which comprises three categories of findings. First, an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences is combined with a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, two of the four candidate strategies identified by this effort are assessed in detail. These are (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

Carry

DEFF Research Database (Denmark)

Koijen, Ralph S.J.; Moskowitz, Tobias J.; Heje Pedersen, Lasse

that include global equities, global bonds, currencies, commodities, US Treasuries, credit, and equity index options. This predictability underlies the strong returns to "carry trades" that go long high-carry and short low-carry securities, applied almost exclusively to currencies, but shown here...

Ageing study of Cirus reactor vessel expansion bellow

International Nuclear Information System (INIS)

Ramana, W.V.; Dutta, B.K.; Kushwaha, H.S.; Sahu, A.K.; Bhatnagar, A.; Pant, R.C.

1994-01-01

Expansion bellow of Cirus reactor vessel is a comparatively weak component which is joined to top tube sheet and shell by helium tight lap weld. This has been subjected to thermal stress caused by high temperature during reactor operation and thermal shock due to trip or shutdown. Therefore a finite element analysis was carried out to assess thermal stresses and fatigue life of the component. It was found that the fluctuating stress in the bellow is far less than its endurance limit. (author). 2 tabs., 3 figs

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

Directory of Open Access Journals (Sweden)

Yuanyuan Nie

2017-09-01

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

Research Vessel R/V Sikuliaq: Joining the UNOLS Fleet in 2014

Science.gov (United States)

Whitledge, T. E.

2013-12-01

The global class research vessel R/V Sikuliaq is being constructed on behalf of the NSF to support future scientific studies in high latitude waters. The 261 foot vessel will be capable of breaking 2.5 foot thick ice at 2 knots with an endurance of 45 days at sea and cruising at 11 knots. The R/V Sikuliaq has a beam of 52 feet and a draft of 18.9 feet that will carry 26 scientists and a crew of 20. Berthing accommodations are a combination of single/double rooms with one stateroom and the common areas of the vessel are designed for ADA access and accommodations. The total laboratory space (main, analytical, electronics, wet, upper, and Baltic room are 2100 square feet. The 4360 square foot working deck that is approximately 70 feet in length will accommodate 2-4 vans and multiple science operations. The vessel design strives to have the lowest possible environmental impact, including a low underwater-radiated noise signature. The science systems are prescribed to be state-of-the-art for bottom mapping, over-the-side 'hands free' gear handling, broad band communications and scientific walk-in freezer and environmental chamber. More details and photos of the construction progress are available on the website at www.sfos.uaf.edu/arrv. The vessel was launched in October 2012 and delivery to the University of Alaska Fairbanks is scheduled for November 2013. Scientific operations following testing and science sea trials are planned to start in summer of 2014. Questions about the science systems or vessel capabilities should be directed to Terry Whitledge (terry@ims.uaf.edu).

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

International Nuclear Information System (INIS)

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

1988-01-01

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

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

Ductile fracture toughness of modified A 302 Grade B Plate materials, data analysis. Volume 1

Energy Technology Data Exchange (ETDEWEB)

McCabe, D.E.; Manneschmidt, E.T.; Swain, R.L.

1997-01-01

The goal of this work was to develop ductile fracture toughness data in the form of J-R curves for modified A302 grade B plate materials typical of those used in reactor pressure vessels. A previous experimental study on one heat of A302 grade B plate showed decreasing J-R curves with increased specimen thickness. This characteristic has not been observed in tests made on recent production materials of A533 grade B and A508 class 2 pressure vessel steels. It was unknown if the departure from norm for the material was a generic characteristic for all heats of A302 grade B steels or unique to that particular plate. Seven heats of modified A302 grade B steel and one heat of vintage A533 grade B steel were tested for chemical content, tensile properties, Charpy transition temperature curves, drop-weight nil-ductility transition (NDT) temperature, and J-R curves. Tensile tests were made in the three principal orientations and at four temperatures, ranging from room temperature to 550F. Charpy V-notch transition temperature curves were obtained in longitudinal, transverse, and short transverse orientations. J-R curves were made using four specimen sizes (1/2T, 1T, 2T, and 4T). The fracture mechanics-based evaluation method covered three test orientations and three test temperatures (80, 400, and 550F). However, the coverage of these variables was contingent upon the amount of material provided. Drop-weight NDT temperature was determined for the T-L orientation only. None of the heats of modified A302 grade B showed size effects of any consequence on the J-R curve behavior. Crack orientation effects were present, but none were severe enough to be reported as atypical. A test temperature increase from 180 to 550F produced the usual loss in J-R curve fracture toughness. Generic J-R curves and curve fits were generated to represent each heat of material. This volume deals with the evaluation of data and the discussion of technical findings. 8 refs., 18 figs., 8 tabs.

Ductile fracture toughness of modified A 302 Grade B Plate materials, data analysis. Volume 1

International Nuclear Information System (INIS)

McCabe, D.E.; Manneschmidt, E.T.; Swain, R.L.

1997-01-01

The goal of this work was to develop ductile fracture toughness data in the form of J-R curves for modified A302 grade B plate materials typical of those used in reactor pressure vessels. A previous experimental study on one heat of A302 grade B plate showed decreasing J-R curves with increased specimen thickness. This characteristic has not been observed in tests made on recent production materials of A533 grade B and A508 class 2 pressure vessel steels. It was unknown if the departure from norm for the material was a generic characteristic for all heats of A302 grade B steels or unique to that particular plate. Seven heats of modified A302 grade B steel and one heat of vintage A533 grade B steel were tested for chemical content, tensile properties, Charpy transition temperature curves, drop-weight nil-ductility transition (NDT) temperature, and J-R curves. Tensile tests were made in the three principal orientations and at four temperatures, ranging from room temperature to 550F. Charpy V-notch transition temperature curves were obtained in longitudinal, transverse, and short transverse orientations. J-R curves were made using four specimen sizes (1/2T, 1T, 2T, and 4T). The fracture mechanics-based evaluation method covered three test orientations and three test temperatures (80, 400, and 550F). However, the coverage of these variables was contingent upon the amount of material provided. Drop-weight NDT temperature was determined for the T-L orientation only. None of the heats of modified A302 grade B showed size effects of any consequence on the J-R curve behavior. Crack orientation effects were present, but none were severe enough to be reported as atypical. A test temperature increase from 180 to 550F produced the usual loss in J-R curve fracture toughness. Generic J-R curves and curve fits were generated to represent each heat of material. This volume deals with the evaluation of data and the discussion of technical findings. 8 refs., 18 figs., 8 tabs

Sense, enrich and classify: The scanmaris workshop for assessment of vessel's abnormal behavior in the EEZ

OpenAIRE

Jangal , Florent; Giraud , Marie-Annick; Morel , Michel; Mano , Jean-Pierre; Napoli , Aldo; Littaye , Anne

2008-01-01

International audience; Constant monitoring of the Exclusive Economic Zone cannot be performed only using high performance sensors. On the one hand, all available information on the observed area as juridical history of vessels or delineation of fishing zone is not necessarily measurable. On the other hand, even if the large amount of available information could be caught out they would be useless if none thorough sorting and analysis are carried on. So, we propose to sense vessel trail in Ex...

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Prospective Mathematics Teachers' Ability to Identify Mistakes Related to Angle Concept of Sixth Grade Students

Science.gov (United States)

Arslan, Cigdem; Erbay, Hatice Nur; Guner, Pinar

2017-01-01

In the present study we try to highlight prospective mathematics teachers' ability to identify mistakes of sixth grade students related to angle concept. And also we examined prospective mathematics teachers' knowledge of angle concept. Study was carried out with 30 sixth-grade students and 38 prospective mathematics teachers. Sixth grade students…

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

The potential role of perivascular lymphatic vessels in preservation of kidney allograft function.

Science.gov (United States)

Tsuchimoto, Akihiro; Nakano, Toshiaki; Hasegawa, Shoko; Masutani, Kosuke; Matsukuma, Yuta; Eriguchi, Masahiro; Nagata, Masaharu; Nishiki, Takehiro; Kitada, Hidehisa; Tanaka, Masao; Kitazono, Takanari; Tsuruya, Kazuhiko

2017-08-01

Lymphangiogenesis occurs in diseased native kidneys and kidney allografts, and correlates with histological injury; however, the clinical significance of lymphatic vessels in kidney allografts is unclear. This study retrospectively reviewed 63 kidney transplant patients who underwent protocol biopsies. Lymphatic vessels were identified by immunohistochemical staining for podoplanin, and were classified according to their location as perivascular or interstitial lymphatic vessels. The associations between perivascular lymphatic density and kidney allograft function and pathological findings were analyzed. There were no significant differences in perivascular lymphatic densities in kidney allograft biopsy specimens obtained at 0 h, 3 months and 12 months. The groups with higher perivascular lymphatic density showed a lower proportion of progression of interstitial fibrosis/tubular atrophy grade from 3 to 12 months (P for trend = 0.039). Perivascular lymphatic density was significantly associated with annual decline of estimated glomerular filtration rate after 12 months (r = -0.31, P = 0.017), even after adjusting for multiple confounders (standardized β = -0.30, P = 0.019). High perivascular lymphatic density is associated with favourable kidney allograft function. The perivascular lymphatic network may be involved in inhibition of allograft fibrosis and stabilization of graft function.

Project management techniques used in the European Vacuum Vessel sectors procurement for ITER

Energy Technology Data Exchange (ETDEWEB)

Losasso, Marcello, E-mail: marcello.losasso@f4e.europa.eu [Fusion for Energy (F4E), Barcelona (Spain); Ortiz de Zuniga, Maria; Jones, Lawrence; Bayon, Angel; Arbogast, Jean-Francois; Caixas, Joan; Fernandez, Jose; Galvan, Stefano; Jover, Teresa [Fusion for Energy (F4E), Barcelona (Spain); Ioki, Kimihiro [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lewczanin, Michal; Mico, Gonzalo; Pacheco, Jose Miguel [Fusion for Energy (F4E), Barcelona (Spain); Preble, Joseph [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Stamos, Vassilis; Trentea, Alexandru [Fusion for Energy (F4E), Barcelona (Spain)

2012-08-15

Highlights: Black-Right-Pointing-Pointer File name contains the directory tree structure with a string of three-letter acronyms, thereby enabling parent directory location when confronted with orphan files. Black-Right-Pointing-Pointer The management of the procurement procedure was carried out in an efficient and timely manner, achieving precisely the contract placement date foreseen at the start of the process. Black-Right-Pointing-Pointer The contract start-up has been effectively implemented and a flexible project management system has been put in place for an efficient monitoring of the contract. - Abstract: The contract for the seven European Sectors of the ITER Vacuum Vessel (VV) was placed at the end of 2010 with a consortium of three Italian companies. The task of placing and the initial take-off of this large and complex contract, one of the largest placed by F4E, the European Domestic Agency for ITER, is described. A stringent quality controlled system with a bespoke Vacuum Vessel Project Lifecycle Management system to control the information flow, based on ENOVIA SmarTeam, was developed to handle the storage and approval of Documentation including links to the F4E Vacuum Vessel system and ITER International Organization System interfaces. The VV Sector design and manufacturing schedule is based on Primavera software, which is cost loaded thus allowing F4E to carry out performance measurement with respect to its payments and commitments. This schedule is then integrated into the overall Vacuum Vessel schedule, which includes ancillary activities such as instruments, preliminary design and analysis. The VV Sector Risk Management included three separate risk analyses from F4E and the bidders, utilizing two different methodologies. These efforts will lead to an efficient and effective implementation of this contract, vital to the success of the ITER machine, since the Vacuum Vessel is the biggest single work package of Europe's contribution to ITER and

Studies and development of essential systems in the surveillance program, life extension potential of the vessel and master curve in nuclear power plants

International Nuclear Information System (INIS)

Romero C, J.; Hernandez C, R.; Rocamontes A, M.; Perez R, N.

2010-01-01

The nuclear power plants owners should demonstrate that the effects of the embrittlement by neutronic radiation do not commit the structural integrity of the pressure vessel of the nuclear reactors, so much under conditions of routine operation as below an accident postulate. In consequence, in Mexico surveillance programs of the vessels of the nuclear power plant of Laguna Verde exist, in which three surveillance capsules are have by reactor. A surveillance capsule is composed by a support and between six and eight containers for test tubes and dosemeters. The containers for test tubes are of two types: rectangular container for Charpy V test tubes and cylindrical container for tension test tubes. These test tubes are subject to a same or bigger neutronic flow that of the vessel, being representative witness of the mechanical conditions of the vessel. The objective of to assay the test tubes to impact is to evaluate the embrittlement grade of the vessel beforehand during its useful life of operation, as well as to determinate the running of the ductile-fragile transition temperature in function of the time. (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

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.

Baking system for ports of experimental advanced super-conducting tokamak vacuum vessel and thermal stress analysis

International Nuclear Information System (INIS)

Cheng Yali; Bao Liman; Song Yuntao; Yao Damao

2006-01-01

The baking system of Experimental Advanced Super-Conducting Toakamk (EAST) vacuum vessel is necessary to obtain the baking temperature of 150 degree C. In order to define suitable alloy heaters and achieve their reasonable layouts, thermal analysis was carried out with ANSYS code. The analysis results indicate that the temperature distribution and thermal stress of most parts of EAST vacuum vessel ports are uniform, satisfied for the requirement, and are safe based on ASME criterion. Feasible idea on reducing the stress focus is also considered. (authors)

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

Materials for high temperature reactor vessels

International Nuclear Information System (INIS)

Buenaventura Pouyfaucon, A.

2004-01-01

Within the 5th Euraton Framework Programme, a big effort is being made to promote and consolidate the development of the High Temperature Reactor (HTR). Empresarios Agrupados is participating in this project and among others, also forms part of the HTR-M project Materials for HTRs. This paper summarises the work carried out by Empresarios Agrupados regarding the material selection of the HTR Reactor Pressure Vessel (RPV). The possible candidate materials and the most promising ones are discussed. Design aspects such as the RPV sensitive zones and material damage mechanisms are considered. Finally, the applicability of the existing design Codes and Standards for the design of the HTR RPV is also discussed. (Author)

Stress Rupture Life Reliability Measures for Composite Overwrapped Pressure Vessels

Science.gov (United States)

Murthy, Pappu L. N.; Thesken, John C.; Phoenix, S. Leigh; Grimes-Ledesma, Lorie

2007-01-01

Composite Overwrapped Pressure Vessels (COPVs) are often used for storing pressurant gases onboard spacecraft. Kevlar (DuPont), glass, carbon and other more recent fibers have all been used as overwraps. Due to the fact that overwraps are subjected to sustained loads for an extended period during a mission, stress rupture failure is a major concern. It is therefore important to ascertain the reliability of these vessels by analysis, since the testing of each flight design cannot be completed on a practical time scale. The present paper examines specifically a Weibull statistics based stress rupture model and considers the various uncertainties associated with the model parameters. The paper also examines several reliability estimate measures that would be of use for the purpose of recertification and for qualifying flight worthiness of these vessels. Specifically, deterministic values for a point estimate, mean estimate and 90/95 percent confidence estimates of the reliability are all examined for a typical flight quality vessel under constant stress. The mean and the 90/95 percent confidence estimates are computed using Monte-Carlo simulation techniques by assuming distribution statistics of model parameters based also on simulation and on the available data, especially the sample sizes represented in the data. The data for the stress rupture model are obtained from the Lawrence Livermore National Laboratories (LLNL) stress rupture testing program, carried out for the past 35 years. Deterministic as well as probabilistic sensitivities are examined.

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)

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

Metallography and microstructure interpretation of some archaeological tin bronze vessels from Iran

Energy Technology Data Exchange (ETDEWEB)

Oudbashi, Omid, E-mail: o.oudbashi@aui.ac.ir [Department of Conservation of Historic Properties, Faculty of Conservation, Art University of Isfahan, Hakim Nezami Street, Sangtarashha Alley, P.O. Box 1744, Isfahan (Iran, Islamic Republic of); Davami, Parviz, E-mail: pdavami@razi-foundation.com [Faculty of Material Science and Engineering, Sharif University of Technology/Razi Applied Science Foundation, No. 27, Fernan St., Shahid Ghasem Asghari Blvd., km 21 of Karadj Makhsous Road, Tehran (Iran, Islamic Republic of)

2014-11-15

Archaeological excavations in western Iran have recently revealed a significant Luristan Bronzes collection from Sangtarashan archaeological site. The site and its bronze collection are dated to Iron Age II/III of western Iran (10th–7th century BC) according to archaeological research. Alloy composition, microstructure and manufacturing technique of some sheet metal vessels are determined to reveal metallurgical processes in western Iran in the first millennium BC. Experimental analyses were carried out using Scanning Electron Microscopy–Energy Dispersive X-ray Spectroscopy and Optical Microscopy/Metallography methods. The results allowed reconstructing the manufacturing process of bronze vessels in Luristan. It proved that the samples have been manufactured with a binary copper–tin alloy with a variable tin content that may relates to the application of an uncontrolled procedure to make bronze alloy (e.g. co-smelting or cementation). The presence of elongated copper sulphide inclusions showed probable use of copper sulphide ores for metal production and smelting. Based on metallographic studies, a cycle of cold working and annealing was used to shape the bronze vessels. - Highlights: • Sangtarashan vessels are made by variable Cu-Sn alloys with some impurities. • Various compositions occurred due to applying uncontrolled smelting methods. • The microstructure represents thermo-mechanical process to shape bronze vessels. • In one case, the annealing didn’t remove the eutectoid remaining from casting. • The characteristics of the bronzes are similar to other Iron Age Luristan Bronzes.

Analysis and evaluation system for elevated temperature design of pressure vessels

International Nuclear Information System (INIS)

Hayakawa, Teiji; Sayawaki, Masaaki; Nishitani, Masahiro; Mii, Tatsuo; Murasawa, Kanji

1977-01-01

In pressure vessel technology, intensive efforts have recently been made to develop the elevated temperature design methods. Much of the impetus of these efforts has been provided mainly by the results of the Liquid Metal Fast Breeder Reactor (LMFBR) and more recently, of the High Temperature Gas-cooled Reactor (HTGR) Programs. The pressure vessels and associated components in these new type nuclear power plants must operate for long periods at elevated temperature where creep effects are significant and then must be designed by rigorous analysis for high reliability and safety. To carry out such an elevated temperature designing, numbers of highly developed analysis and evaluation techniques, which are so complicated as to be impossible by manual work, are indispensable. Under these circumstances, the authors have made the following approaches in the study: (1) Study into basic concepts and the associated techniques in elevated temperature design. (2) Systematization (Analysis System) of the procedure for loads and stress analyses. (3) Development of post-processor, ''POST-1592'', for strength evaluation based on ASME Code Case 1592-7. By linking the POST-1592 together with the Analysis System, an analysis and evaluation system is developed for an elevated temperature design of pressure vessels. Consequently, designing of elevated temperature vessels by detailed analysis and evaluation has easily and effectively become feasible by applying this software system. (auth.)

Design of the Intersector Welding Robot for vacuum vessel assembly and maintenance

International Nuclear Information System (INIS)

Jones, L.; Dagenais, J.-F.; Daenner, W.; Maisonnier, D.

2000-01-01

Next Step Fusion Devices require on-site (field weld) joining of sectors of the thick-walled vacuum vessel for structural and vacuum integrity. EFDA (European Fusion Development Agreement) is supporting an R and D programme to investigate processes for assembly of the vacuum vessel and to carry out cutting, re-welding and inspection for remote sector replacement, forming part of the overall VV/blanket research effort. In order to direct the process end-effectors along the field joint zone, a track-mounted Intersector Welding Robot (IWR) on a mock-up of a region of the vacuum vessel has been designed and is described in this paper. A rail-mounted hexapod type robot offers six axes of motion over a limited work envelope with high payload to robot weight ratio. A solution to the production of reduced pressure local vacuum is the installation of short, lightweight segments bolted to each other and the vessel wall. The various process heads can be mounted using end-effectors of special design. To minimise the supply and interface problems for the IWR prototype, its motion control and electronic systems will be embedded locally. A laser scan with camera forms the on-line seam tracking capability to compensate for rail and seam deviations

Fabrication history and mechanical properties for ASTM A-533 Grade-B Class-2 steel weld for fully welded nuclear pressure vessels

International Nuclear Information System (INIS)

Pachur, D.

1979-01-01

Till now pressure vessels for light water reactors were made from rolled plates and forgings connected with each other by welding. The optimal quality of plates and forgings are limited in principle by the foundry technology. It is well known that in this process decomposition and segregation zones occur. Besides the heat affected zone created by the welding process is a weak link. The heat affected zone is heterogeneous and can be harbinger of risks leading to cracks. The production of a pressure vessel through shape welding is an alternative. The cylindrical container is produced by the application of one layer of welding after the other in a preshaped form. During the welding process the previously applied layers are simultaneously being tempered. The undesirable chemical residual elements are evenly distributed and segregation zones do not occur. Since we have only welding material the disadvantages of a heat affected zone are avoided. Furthermore the mechanical properties are independent of location and orientation. This shape welding process proved to be highly economical already during the experimental stay. Besides this process is applicable for vessel of any desired dimension

The elevated temperature and thermal shock fracture toughnesses of nuclear pressure vessel steel

International Nuclear Information System (INIS)

Hirano, Kazumi; Kobayashi, Hideo; Nakazawa, Hajime; Nara, Atsushi.

1979-01-01

Thermal shock experiments were conducted on nuclear pressure vessel steel A533 Grade B Class 1. Elastic-plastic fracture toughness tests were carried out within the same high temperature range of the thermal shock experiment and the relation between stretched zone width, SZW and J-integral was clarified. An elastic-plastic thermal shock fracture toughness value. J sub(tsc) was evaluated from a critical value of stretched zone width, SZW sub(tsc) at the initiation of thermal shock fracture by using the relation between SZW and J. The J sub(tsc) value was compared with elastic-plastic fracture toughness values, J sub( ic), and the difference between the J sub(tsc) and J sub( ic) values was discussed. The results obtained are summarized as follows; (1) The relation between SZW and J before the initiation of stable crack growth in fracture toughness test at a high temperature can be expressed by the following equation regardless of test temperature, SZW = 95(J/E), where E is Young's modulus. (2) Elevated temperature fracture toughness values ranging from room temperature to 400 0 C are nearly constant regardless of test temperature. It is confirmed that upper shelf fracture toughness exists. (3) Thermal shock fracture toughness is smaller than elevated temperature fracture toughness within the same high temperature range of thermal shock experiment. (author)

Equipment Qualification and Environment Establishment for COTS Dedication of Safety Grade PC

International Nuclear Information System (INIS)

Ahn, Kwang Chul; Bin, Chang Sun; Kwon, Yoon Kwang; Kang, Shin Woo; Koh, Sung Won; Shon, Eui Chan; Jang, Hyun Doo; Song, Il Sup; Lee, Hyun Noh

2010-08-01

Safety grade PCs are required for protection systems of SKN 3 and 4 nuclear power plant and subsequent plants. Commercial grade item (CGI) dedication should be carried out to utilize a commercial PC as a safety grade PC of nuclear power plants. This project is aimed to perform the equipment qualification of the commercial PC, and review and improve the quality system of the PC supplier. As a result of the EQ for the CGI dedication, selected military PCs have passed the environment test, the seismic test, and the EMI test required for the digital controllers of nuclear plants. In addition, a walk-through of the quality system of the PC supplier, ISO 9001, was carried out and the quality system was improved. History data for the PC was gathered. As the analysis of the history data showed that operating experience time of the PC is longer than the plant life time, the history data could be used as an evidence of acceptance. After dedicated according to the CGI dedication process, military rugged PCs could be used for safety grade PCs

[Hemodynamic phenomena in retrobulhar and eyeball vessels].

Science.gov (United States)

Modrzejewska, Monika

2011-01-01

The purpose of this review was to evaluate factors connected with blood flow and indices regulating vascular diameter and some parameters influencing retrobulbar circulation such as type of vascular resistance, anatomical structure of vascular wall and vessel lumen. Neurogenic and angiogenic factors, rheological blood composition, presence of anatomical and pathological obstructions on blood flow pathway as well as degree of development of collateral circulation pathways--have influence on the volume and blood flow velocity in eyeball. There were discussed bulbar circulation hemodynamics, emphasizing the importance of perfusion pressure. The role of risk factors was underlined for pathological lesions in vessels supplying blood to eyeball and in ophthalmic artery (OA) and its collaterals, in central retinal artery (CRA) as well as posterior ciliary arteries (PCAs), and in venous system carrying away blood from eye. IN CONCLUSION--the results of many studies of retrobulbar blood flow in different types of ophthalmic diseases of the vascular etiopathogenesis indicate that registry of the mean values of blood flow parameters and vascular resistance indices parallel to measurement of blood flow spectrum in OA, CRA, PCAs arteries, might contribute much information to explain or to evaluate nature of pathological changes in retinal and choroidal circulation.

Annealing the reactor vessel at the Palisades Plant

International Nuclear Information System (INIS)

Fenech, R.A.

1996-01-01

In the way of background, Palisades was licensed in 1967 and went commercial in 1971. Jumping to two years ago, we faced at that time three issues that challenged our ability to operate to end-of-license, which would be 2007 without any extensions. The three items were regulatory performance, economic performance, and reactor vessel embrittlement. We had not been operating the plant with the kind of conservative decisions and with the kind of safety margins that one is expected to operate a plant in the United States at this time. Our economic performance was not satisfactory in that our capacity factor was low and our costs high. In the area of reactor vessel embrittlement, our analysis showed that we would reach the NRC screening criteria for embrittlement in the year 2004. Over the last two years, we have made significant improvements in the first two areas. Our decision-making has changed. Our performance, especially over the last year and a half, has been excellent. In addition, we have gotten our capacity factors up and our costs under control. Clearly, sustained performance is what is going to carry the day but from what we can see and from where we are, we are in more of a maintenance-of-performance than in a turn-around situation. On the other hand, in the area of reactor vessel embrittlement, about a year and a half ago we had a bit of a setback. We had taken material from retired steam generators that had welds identical to the welds in our reactor vessel. When we analyzed the welds from our steam generators, we were given some surprises about the chemistry makeup. When we applied the new information to our analysis, we changed the date on which we would reach our screening criteria from 2004 to late 1999

Investigation of the design of a metal-lined fully wrapped composite vessel under high internal pressure

Science.gov (United States)

Kalaycıoğlu, Barış; Husnu Dirikolu, M.

2010-09-01

In this study, a Type III composite pressure vessel (ISO 11439:2000) loaded with high internal pressure is investigated in terms of the effect of the orientation of the element coordinate system while simulating the continuous variation of the fibre angle, the effect of symmetric and non-symmetric composite wall stacking sequences, and lastly, a stacking sequence evaluation for reducing the cylindrical section-end cap transition region stress concentration. The research was performed using an Ansys® model with 2.9 l volume, 6061 T6 aluminium liner/Kevlar® 49-Epoxy vessel material, and a service internal pressure loading of 22 MPa. The results show that symmetric stacking sequences give higher burst pressures by up to 15%. Stacking sequence evaluations provided a further 7% pressure-carrying capacity as well as reduced stress concentration in the transition region. Finally, the Type III vessel under consideration provides a 45% lighter construction as compared with an all metal (Type I) vessel.

Optimization and studies of the welding processes, automation of the sealing welding system and fracture mechanics in the vessels surveillance in nuclear power plants

International Nuclear Information System (INIS)

Gama R, G.

2011-01-01

Inside this work the optimization of two welding systems is described, as well as the conclusion of a system for the qualification of containers sealing in the National Institute of Nuclear Research that have application in the surveillance programs of nuclear reactors vessels and the correspondent extension of the operation license. The test tubes Charpy are assay to evaluate the embrittlement grade, when obtaining the increment in the reference temperature and the decrease of the absorbed maximum energy, in the transition curve fragile-ductile of the material. After the test two test tube halves are obtained that should take advantage to follow the surveillance of the vessel and their possible operation extension, this is achieved by means of rebuilding (being obtained of a tested test tube two reconstituted test tubes). The welding system for the rebuilding of test tubes Charpy, was optimized when diminishing the union force at solder, achieving the elimination of the rejection for penetration lack for spill. For this work temperature measurements were carried out at different distances of the welding interface from 1 up to 12 mm, obtaining temperature profiles. With the maximum temperatures were obtained a graph and equation that represents the maximum temperature regarding the distance of the interface, giving as a result practical the elimination of other temperature measurements. The reconstituted test tubes were introduced inside pressurized containers with helium of ultra high purity to 1 pressure atmosphere. This process was carried out in the welding system for containers sealing, where an automatic process was implemented by means of an application developed in the program LabVIEW, reducing operation times and allowing the remote control of the process, the acquisition parameters as well as the generation of welding reports, avoiding with this the human error. (Author)

Execution of programme of post-service study of the condition of nuclear icebreaker Lenin reactor 1 pressure vessel metal and perspectives of application of results to increase service life of nuclear icebreakers reactor vessels

International Nuclear Information System (INIS)

Platonov, P.Ya.; Shtrombakh, Ya.I.; Amaev, A.D.; Krasikov, E.A.; Korolev, Yu.N.; Zabusov, O.O.; Glushakov, G.M.

2001-01-01

With the aim of determining the irradiation-induced embrittlement of a base metal and a weld metal in a pressure vessel of the nuclear icebreaker Lenin after 18 years operation the specimens cut out of a vessel wall are used to study the chemical composition and to carry out impact tests. From the test results the temperature dependences of fracture energy are built which define the irradiation embrittlement of a low alloy steel. It is noted that the annealing at 475 deg C for 100 h results in complete restoration of impact strength. Based on the results obtained the following conclusions are formulated: a reactor vessel base metal has high resistance to brittle fracture and high radiation resistance; a weld metal possesses rather high radiation resistance but unsatisfactory ductile-brittle transition temperature (∼ 63 deg C); for cladded vessels there is a potential reserve in the form of enhanced radiation resistance of an undercladding layer; in the final stage of operation the coolant temperature is recommended to be kept at the highest possible level [ru

Vascular targeting of LIGHT normalizes blood vessels in primary brain cancer and induces intratumoural high endothelial venules.

Science.gov (United States)

He, Bo; Jabouille, Arnaud; Steri, Veronica; Johansson-Percival, Anna; Michael, Iacovos P; Kotamraju, Venkata Ramana; Junckerstorff, Reimar; Nowak, Anna K; Hamzah, Juliana; Lee, Gabriel; Bergers, Gabriele; Ganss, Ruth

2018-06-01

High-grade brain cancer such as glioblastoma (GBM) remains an incurable disease. A common feature of GBM is the angiogenic vasculature, which can be targeted with selected peptides for payload delivery. We assessed the ability of micelle-tagged, vascular homing peptides RGR, CGKRK and NGR to specifically bind to blood vessels in syngeneic orthotopic GBM models. By using the peptide CGKRK to deliver the tumour necrosis factor (TNF) superfamily member LIGHT (also known as TNF superfamily member 14; TNFSF14) to angiogenic tumour vessels, we have generated a reagent that normalizes the brain cancer vasculature by inducing pericyte contractility and re-establishing endothelial barrier integrity. LIGHT-mediated vascular remodelling also activates endothelia and induces intratumoural high endothelial venules (HEVs), which are specialized blood vessels for lymphocyte infiltration. Combining CGKRK-LIGHT with anti-vascular endothelial growth factor and checkpoint blockade amplified HEV frequency and T-cell accumulation in GBM, which is often sparsely infiltrated by immune effector cells, and reduced tumour burden. Furthermore, CGKRK and RGR peptides strongly bound to blood vessels in freshly resected human GBM, demonstrating shared peptide-binding activities in mouse and human primary brain tumour vessels. Thus, peptide-mediated LIGHT targeting is a highly translatable approach in primary brain cancer to reduce vascular leakiness and enhance immunotherapy. Copyright © 2018 Pathological Society of Great Britain and Ireland. Published by John Wiley & Sons, Ltd. Copyright © 2018 Pathological Society of Great Britain and Ireland. Published by John Wiley & Sons, Ltd.

Development of containers sealing system like part of surveillance program of the vessel in nuclear power plants

International Nuclear Information System (INIS)

Romero C, J.; Hernandez C, R.; Fernandez T, F.; Rocamontes A, M.; Perez R, N.

2009-10-01

The owners of nuclear power plants should be demonstrate that the embrittlement effects by neutronic radiation do not commit the structural integrity from the pressure vessel of nuclear reactors, during conditions of routine operation and below postulate accident. For this reason, there are surveillance programs of vessels of nuclear power plants, in which are present surveillance capsules. A surveillance capsule is compound by the support, six containers for test tubes and dosimeters. The containers for test tubes are of two types: rectangular container for test tubes, Charpy V and Cylindrical Container for tension test tubes. These test tubes are subject to a same or bigger neutronic flow to that of vessel, being representative of vessel mechanical conditions. The test tubes are rehearsed to watch over the increase of embrittlement that presents the vessel. This work describes the development of welding system to seal the containers for test tubes, these should be filled with helium of ultra high purity, to a pressure of an atmosphere. In this system the welding process Gas Tungsten Arc Welding is used, a hermetic camera that allows to place the containers with three grades of freedom, a vacuum subsystem and pressure, high technology equipment's like: power source with integrated computer, arc starter of high frequency, helium flow controller, among others. Finally, the advances in the inspection system for the qualification of sealing system are mentioned, system that should measure the internal pressure of containers and the helium purity inside these. (Author)

Ultrasonic inspection of the Calder Hall and Chaplecross reactor pressure vessels

International Nuclear Information System (INIS)

Pennick, A.M.

1993-01-01

This paper describes the ultrasonic inspection surveys that have recently been carried out on the Calder Hall and Chapelcross Magnox steel reactor pressure vessels. The development of the inspection system, which is based on the Rediman manipulator and uses the Sonomatic Zipscan equipment and Time-of-Flight diffraction techniques is discussed. The inspection results are presented and compared with the original inspection findings and limiting crack sizes. (author)

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.

A simple method for assigning genomic grade to individual breast tumours

International Nuclear Information System (INIS)

Wennmalm, Kristian; Bergh, Jonas

2011-01-01

The prognostic value of grading in breast cancer can be increased with microarray technology, but proposed strategies are disadvantaged by the use of specific training data or parallel microscopic grading. Here, we investigate the performance of a method that uses no information outside the breast profile of interest. In 251 profiled tumours we optimised a method that achieves grading by comparing rank means for genes predictive of high and low grade biology; a simpler method that allows for truly independent estimation of accuracy. Validation was carried out in 594 patients derived from several independent data sets. We found that accuracy was good: for low grade (G1) tumors 83- 94%, for high grade (G3) tumors 74- 100%. In keeping with aim of improved grading, two groups of intermediate grade (G2) cancers with significantly different outcome could be discriminated. This validates the concept of microarray-based grading in breast cancer, and provides a more practical method to achieve it. A simple R script for grading is available in an additional file. Clinical implementation could achieve better estimation of recurrence risk for 40 to 50% of breast cancer patients

A simple method for assigning genomic grade to individual breast tumours

Directory of Open Access Journals (Sweden)

Bergh Jonas

2011-07-01

Full Text Available Abstract Background The prognostic value of grading in breast cancer can be increased with microarray technology, but proposed strategies are disadvantaged by the use of specific training data or parallel microscopic grading. Here, we investigate the performance of a method that uses no information outside the breast profile of interest. Results In 251 profiled tumours we optimised a method that achieves grading by comparing rank means for genes predictive of high and low grade biology; a simpler method that allows for truly independent estimation of accuracy. Validation was carried out in 594 patients derived from several independent data sets. We found that accuracy was good: for low grade (G1 tumors 83- 94%, for high grade (G3 tumors 74- 100%. In keeping with aim of improved grading, two groups of intermediate grade (G2 cancers with significantly different outcome could be discriminated. Conclusion This validates the concept of microarray-based grading in breast cancer, and provides a more practical method to achieve it. A simple R script for grading is available in an additional file. Clinical implementation could achieve better estimation of recurrence risk for 40 to 50% of breast cancer patients.

Device for the simultaneous operation of the closing valve of a vessel and the closing valve of a transport container

International Nuclear Information System (INIS)

Tellier, Claude; Surriray, Michel.

1982-01-01

This device includes mechanisms for unlatching the closing valve of the vessel and securing it to the closing valve of the transport container and other mechanisms for vertically raising the assembly of valves, pivoting it and bringing it into a vertical position in a bulge provided in the bottom of the transport container. For example the first containment is a nuclear reactor vessel and the transport container is used for carrying an item from the vessel to an external area (for instance, a defective pump to the repair area) and for the return transport operation [fr

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

Development of PIE techniques for irradiated LWR pressure vessel steels

International Nuclear Information System (INIS)

Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide

1999-01-01

For the evaluation of safety and integrity of light water reactors (LWRs), various post irradiation examinations (PIEs) of reactor pressure vessel (RPV) steels and fuel claddings have been carried out in the Research Hot Laboratory (RHL). In recent years, the instrumented Charpy impact testing machine was remodeled aiming at the improvement of accuracy and reliability. By this remodeling, absorbed energy and other useful information on impact properties can be delivered from the force-displacement curve for the evaluation of neutron irradiation embrittlement behavior of LWR-RPV steels at one-time striking. In addition, two advanced PIE technologies are now under development. One is the remote machining of mechanical test pieces from actual irradiated pressure vessel steels. The other is development of low-cycle and high-cycle fatigue test technology in order to clarify the post-irradiation fatigue characteristics of structural and fuel cladding materials. (author)

Dosimetry of VVER-1000 Reactor Pressure Vessel and Surveillance Specimens as a part of PLiM at Ukrainian NPPs

International Nuclear Information System (INIS)

Bukanov, V.N.; Diemokhin, V.L.; Grytsenko, O.V.; Ilkovych, V.V.; Pugach, A.M.; Pugach, S.M.; Vasylieva, O.G.; Vyshnevskyi, I.M.; Kasatkin, O.G.

2012-01-01

A regular surveillance program for VVER-1000 and its shortages are described. The Methodology for determination of neutron flux functionals on surveillance specimens of VVER-1000 pressure vessel is presented. The radiation exposure monitoring system for VVER-1000 pressure vessel is described. The main principles of an additional surveillance program for VVER-1000 are presented. The Dosimetry Experiment, which is already carrying out at Unit 3 of Rivne NPP, is described. (author)

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

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

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.

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)

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)

Analysis of the in-vessel phase of SAM strategy for a Korean 1000 MWe PWR

Energy Technology Data Exchange (ETDEWEB)

Cho, Sung-Min; Oh, Seung-Jong [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Diab, Aya [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Ain Shams Univ., Cairo (Egypt). Mechanical Power Engineering Dept.

2017-12-15

This paper focuses on the in-vessel phase of Severe Accident Management (SAM) strategy for a Korean 1000 MWe Pressurized Water Reactor (PWR) with reference to ROAAM+ framework approach. To apply ROAAM+, it is needed to identify epistemic and aleatory uncertainties. The selected scenario is a station blackout (SBO) and the corresponding SAM strategy is RCS depressurization followed by water injection into the reactor pressure vessel (RPV). The analysis considers the depressurization timing and the flow rate and timing of in-vessel injection for scenario variations. For the phenomenological uncertainties, the core melting and relocation process is considered to be the most important phenomenon in the in-vessel phase of SAM strategy. Accordingly, a sensitivity analysis is carried out to assess the impact of the cut-off porosity below which the flow area of a core node is zero (EPSCUT), and the critical temperature for cladding rupture (TCLMAX) on the core melting and relocation process. In this paper, the SAM strategy for maintaining the integrity of RPV is derived after quantification of the scenario and phenomenological uncertainties.

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

International Nuclear Information System (INIS)

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

1978-01-01

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

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.)

Evaluation of the transgranular cracking phenomenon on the Indian Point No. 3 steam generator vessels

International Nuclear Information System (INIS)

Czajkowski, C.J.

1985-01-01

A metallurgical investigation was performed on specimens from the shell of steam generators Nos. 31 and 32 of the Indian Point-3 Power Plant. The shell material exhibited high values in hardness which was indicative that relatively high residual stresses may have been present. All observed cracks were transgranular in appearance and were associated with pits on the vessels' inside surfaces. Both stress relieved and non-stress relieved specimens of SA302 Grade B material were tested in a constant extension rate apparatus in various environments in order to reproduce the transgranular cracking at Indian Point No. 3. The paper concludes that SA302 Grade B material is susceptible to transgranular stress corrosion cracking (SCC) in constant extension rate testing (CERT) with as little as 1 ppM chloride (as CUCL 2 ) in 268 0 C H 2 O. 2 refs., 9 figs

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.

Coefficients of Propeller-hull Interaction in Propulsion System of Inland Waterway Vessels with Stern Tunnels

Directory of Open Access Journals (Sweden)

Jan Kulczyk

2014-09-01

Full Text Available Propeller-hull interaction coefficients - the wake fraction and the thrust deduction factor - play significant role in design of propulsion system of a ship. In the case of inland waterway vessels the reliable method of predicting these coefficients in early design stage is missing. Based on the outcomes from model tests and from numerical computations the present authors show that it is difficult to determine uniquely the trends in change of wake fraction and thrust deduction factor resulting from the changes of hull form or operating conditions. Nowadays the resistance and propulsion model tests of inland waterway vessels are carried out rarely because of relatively high costs. On the other hand, the degree of development of computational methods enables’ to estimate the reliable values o interaction coefficients. The computations referred to in the present paper were carried out using the authors’ own software HPSDKS and the commercial software Ansys Fluent.

Containment vessel stability analysis

International Nuclear Information System (INIS)

Harstead, G.A.; Morris, N.F.; Unsal, A.I.

1983-01-01

The stability analysis for a steel containment shell is presented herein. The containment is a freestanding shell consisting of a vertical cylinder with a hemispherical dome. It is stiffened by large ring stiffeners and relatively small longitudinal stiffeners. The containment vessel is subjected to both static and dynamic loads which can cause buckling. These loads must be combined prior to their use in a stability analysis. The buckling loads were computed with the aid of the ASME Code case N-284 used in conjunction with general purpose computer codes and in-house programs. The equations contained in the Code case were used to compute the knockdown factors due to shell imperfections. After these knockdown factors were applied to the critical stress states determined by freezing the maximum dynamic stresses and combining them with other static stresses, a linear bifurcation analysis was carried out with the aid of the BOSOR4 program. Since the containment shell contained large penetrations, the Code case had to be supplemented by a local buckling analysis of the shell area surrounding the largest penetration. This analysis was carried out with the aid of the NASTRAN program. Although the factor of safety against buckling obtained in this analysis was satisfactory, it is claimed that the use of the Code case knockdown factors are unduly conservative when applied to the analysis of buckling around penetrations. (orig.)

An assessment of acoustic emission for nuclear pressure vessel monitoring

International Nuclear Information System (INIS)

Scruby, C.B.

1983-01-01

Recent research has greatly improved our understanding of the basic mechanisms of deformation and fracture that generate detectable acoustic emission signals in structural steels. A critical review of the application of acoustic emission (AE) to the fabrication, proof testing and in-service monitoring of nuclear pressure vessels is presented in the light of this improved understanding. The detectability of deformation and fracture processes in pressure vessel steels is discussed, and recommendations made for improving source location accuracy and the development of quantitative source assessment techniques. Published data suggest that AE can make an important contribution to fabrication monitoring, and to the detection of defects in lower toughness materials during vessel proof testing. In high toughness materials, however, the signals generated during ductile crack growth may frequently be too weak for reliable detection. The feasibility of AE for continuous monitoring has not yet been adequately demonstrated because of high background noise levels and uncertainty about AE signal strengths from the defect growth processes that occur in service. In-service leak detection by AE shows considerable promise. It is recommended that further tests are carried out with realistic defects, and under realistic conditions of loading (including thermal shock and fatigue) and of environment. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1988-12-31

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

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

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.)

French nuclear plants PWR vessel integrity assessment and life management

Energy Technology Data Exchange (ETDEWEB)

Bezdikian, G. [Electricite de France (EDF), Div. Production Nucleaire, 93 - Saint-Denis (France); Quinot, P. [FRAMATOME, Dept. Bloc Reacteur et Boucles Primaires, 92 - Paris-La-Defence (France); Faidy, C.; Churier-Bossennec, H. [Electricite de France (EDF), Div. Ingenierie et Service, 69 - Villeurbanne (France)

2001-07-01

The Reactor Pressure Vessel life management of 56 PWR 3 loop and 4 loop reactors units was engaged by the French Utility EDF (Electricite de France) a few years ago and is yet on going on. This paper will present the work carried out within the framework of justifying why the 34 three loop reactor vessels will remain acceptable for operation for a lifetime of at least 40-years. A summary of the measures will be given. An overall review of actions will be presented describing the French approach, using important existing databases, including studies related to irradiation surveillance monitoring program and end of life fluence assessment. The last results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions) until the end of lifetime, postulating circumferential an radial kinds of flaw located in the stainless steel cladding or shallow sub-cladding area. The results of structural integrity analyses beginning with elastic computations and completed with three-dimensional finite element elastic plastic computations for envelope cases, are compared with code criteria for operating plants. The objective is to evaluate the margins on different parameters as RTNDT (Reference Nil Ductility Transition Temperature), toughness or crack size, to justify the global fitness for service of all these Reactor Pressure Vessels. The paper introduces EDF's maintenance strategy, related to integrity assessment, for those nuclear power plants under operation, based on NDE in-service inspection of the first thirty millimeters in the thickness of the wall and major surveillance programs of the vessels. (author)

French nuclear plants PWR vessel integrity assessment and life management

International Nuclear Information System (INIS)

Bezdikian, G.; Quinot, P.; Faidy, C.; Churier-Bossennec, H.

2001-01-01

The Reactor Pressure Vessel life management of 56 PWR 3 loop and 4 loop reactors units was engaged by the French Utility EDF (Electricite de France) a few years ago and is yet on going on. This paper will present the work carried out within the framework of justifying why the 34 three loop reactor vessels will remain acceptable for operation for a lifetime of at least 40-years. A summary of the measures will be given. An overall review of actions will be presented describing the French approach, using important existing databases, including studies related to irradiation surveillance monitoring program and end of life fluence assessment. The last results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions) until the end of lifetime, postulating circumferential an radial kinds of flaw located in the stainless steel cladding or shallow sub-cladding area. The results of structural integrity analyses beginning with elastic computations and completed with three-dimensional finite element elastic plastic computations for envelope cases, are compared with code criteria for operating plants. The objective is to evaluate the margins on different parameters as RTNDT (Reference Nil Ductility Transition Temperature), toughness or crack size, to justify the global fitness for service of all these Reactor Pressure Vessels. The paper introduces EDF's maintenance strategy, related to integrity assessment, for those nuclear power plants under operation, based on NDE in-service inspection of the first thirty millimeters in the thickness of the wall and major surveillance programs of the vessels. (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

Incremental value of regional wall motion analysis immediately after exercise for the detection of single-vessel coronary artery disease. Study by separate acquisition, dual-isotope ECG-gated single-photon emission computed tomography

International Nuclear Information System (INIS)

Yoda, Shunichi; Sato, Yuichi; Matsumoto, Naoya; Tani, Shigemasa; Takayama, Tadateru; Uchiyama, Takahisa; Saito, Satoshi

2005-01-01

Although the detection of wall motion abnormalities gives incremental value to myocardial perfusion single-photon emission computed tomography (SPECT) in the diagnosis of extensive coronary artery disease (CAD) and high-grade single-vessel CAD, whether or not it is useful in the diagnosis of mild, single-vessel CAD has not been studied previously. Separate acquisition, dual isotope electrocardiogram (ECG)-gated SPECT was performed in 97 patients with a low likelihood of CAD (Group 1) and 46 patients with single-vessel CAD (Group 2). Mild CAD was defined by stenosis of 50-75% (Group 2a, n=22) and moderate to severe CAD was defined by stenosis ≥76% (Group 2b, n=24). Myocardial perfusion and wall motion were graded by a 5 point-scale, 20-segment model. The sensitivity of myocardial perfusion alone was 50% for Group 2a, 83% for Group 2b and 67% for Group 2 as a whole. The overall specificity was 90%. When the wall motion analysis was combined, the sensitivity was increased to 82% in Group 2a and 92% in Group 2b. The ability to detect a wall motion abnormality immediately after exercise gives incremental diagnostic value to myocardial perfusion SPECT in the identification of mild, single-vessel CAD. (author)

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.)

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

International Nuclear Information System (INIS)

Kitagawa, Hideo; Hisada, Toshiaki

1979-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

Penetration of the consolidant Paraloid® B-72 in Macuxi indigenous ceramic vessels investigated by neutron tomography

Science.gov (United States)

Stanojev Pereira, Marco A.; Pugliesi, Reynaldo

2018-05-01

The neutron tomography technique was applied in studying the penetration of the consolidant Paraloid® B-72 in contemporary indigenous ceramic vessels. The study was carried out for two distinct and controlled air humidity conditions, 40% and 90%, in which the vessels were exposed, before the consolidant application. The obtained images have proved that the penetration of Paraloid® B-72 in the ceramic does not depend on the humidity condition in which it was applied, moreover allowed a macro-visualization of the consolidant penetration in the ceramic vessel. As the vessels used in the present work were manufactured by an indigenous artisan, Macuxi, according to the same procedures and raw materials used by the ancient artisans, the results obtained can be used as a guide to assist experts, both in the study of archeological objects of Macuxi origin, as well as other objects that had been made by other tribes that lived in the same Amazon region, in Brazil.

Structural materials for ITER in-vessel component design

Energy Technology Data Exchange (ETDEWEB)

Kalinin, G. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Gauster, W. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Matera, R. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Tavassoli, A.-A.F. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France); Rowcliffe, A. [Oak Ridge National Lab., TN (United States); Fabritsiev, S. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Kawamura, H. [JAERI, IMTR Project, Ibaraki (Japan). Blanket Irradiation Lab.

1996-10-01

The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m{sup 2} in the basic performance phase (BPP)) within a temperature range from 20 to 300 C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350 C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. (orig.).

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...

Applicability of JIS SPV 50 steel to primary containment vessel of nuclear power station

International Nuclear Information System (INIS)

Iida, Kunihiro; Ishikawa, Koji; Sakai, Keiichi; Onozuka, Masakazu; Sato, Makoto.

1979-01-01

The space within reactor containment vessels must be expanded in order to improve the reliability of nuclear power plants, accordingly the adoption of large reactor containment vessels is investigated. SGV 42 and 49 steels in JIS G 3118 have been used for containment vessels so far, but stress relief annealing is required when the thickness exceeds 38 mm. The time has come when the use of thicker conventional plates without stress relieving or the use of high strength steel must be examined in detail. In this study, the tests of confirming material properties were carried out on SPV 50 in JIS G 3115, Steels for pressure vessels, aiming at the method of fabrication without stress relieving. The highest and lowest temperatures in use were set at 171 deg and -8 deg C, respectively. The chemical composition and the mechanical properties of the plates tested, the method of welding, the results of tensile test on the parent metal and the welds, the required lowest preheating temperature, the fracture toughness at low temperature and the brittle fracture causing test are reported. The parent metal and the welded joints of SPV 50 have the properties suitable to reactor containment vessels, namely the sufficient fracture toughness to guarantee the prevention of unstable fracture when the method of welding without stress relieving is adopted. (Kako, I.)

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.)

Prestressed-concrete pressure vessels and their applicability to advanced-energy-system concepts

International Nuclear Information System (INIS)

Naus, D.J.

1983-01-01

Prestressed concrete pressure vessels (PCPVs) are, in essence, spaced steel structures since their strength is derived from a multitude of steel elements made up of deformed reinforcing bars and prestressing tendons which are present in sufficient quantities to carry tension loads imposed on the vessel. Other major components of a PCPV include the concrete, liner and cooling system, and insulation. PCPVs exhibit a number of advantages which make them ideally suited for application to advanced energy concepts: fabricability in virtually any size and shape using available technology, improved safety, reduced capital costs, and a history of proven performance. PCPVs have many applications to both nuclear- and non-nuclear-based energy systems concepts. Several of these concepts will be discussed as well as the research and development activities conducted at ORNL in support of PCPV development

Fabrication and characterization of electrospun poly-L-lactide/gelatin graded tubular scaffolds: Toward a new design for performance enhancement in vascular tissue engineering

Directory of Open Access Journals (Sweden)

A. Yazdanpanah

2015-10-01

Full Text Available In this study, a new design of graded tubular scaffolds have been developed for the performance enhancement in vascular tissue engineering. The graded poly-L-lactide (PLLA and gelatin fibrous scaffolds produced by electrospining were then characterized. The morphology, degradability, porosity, pore size and mechanical properties of four tubular scaffolds (graded PLLA/gelatin, layered PLLA/gelatin, PLLA and gelatin scaffolds have been investigated. The tensile tests demonstrated that the mechanical strength and also the estimated burst pressure of the graded scaffolds were significantly increased in comparison with the layered and gelatin scaffolds. This new design, resulting in an increase in the mechanical properties, suggested the widespread use of these scaffolds in vascular tissue engineering in order to prepare more strengthened vessels.

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.

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.

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

Dynamic fracture toughness of ASME SA508 Class 2a ASME SA533 grade A Class 2 base and heat affected zone material and applicable weld metals

International Nuclear Information System (INIS)

Logsdon, W.A.; Begley, J.A.; Gottshall, C.L.

1978-03-01

The ASME Boiler and Pressure Vessel Code, Section III, Article G-2000, requires that dynamic fracture toughness data be developed for materials with specified minimum yield strengths greater than 50 ksi to provide verification and utilization of the ASME specified minimum reference toughness K/sub IR/ curve. In order to qualify ASME SA508 Class 2a and ASME SA533 Grade A Class 2 pressure vessel steels (minimum yield strengths equal 65 kip/in. 2 and 70 kip/in. 2 , respectively) per this requirement, dynamic fracture toughness tests were performed on these materials. All dynamic fracture toughness values of SA508 Class 2a base and HAZ material, SA533 Grade A Class 2 base and HAZ material, and applicable weld metals exceeded the ASME specified minimum reference toughness K/sub IR/ curve

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...

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

International Nuclear Information System (INIS)

Chatterjee, S.; Shah, Priti Kotak

2008-05-01

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

Mechanical impacts of poloidal eddy currents on the continuous vacuum vessel of a tokamak

International Nuclear Information System (INIS)

In, Sang Ryul; Yoon, Byung Joo.

1996-11-01

Poloidal eddy currents are induced on the continuous torus vacuum vessel by changes of the toroidal field during the machine start-up (toroidal field coil charge), shut-down (toroidal field coil discharge) and plasma disruption (plasma diamagnetism change). Analytic forms for the eddy currents flowing on the vessel, consequent pressures and forces acting on it are presented in this report. The results are applied to typical operation modes of the KT-2 tokamak. Stress analysis for two typical operation modes of toroidal field damping during a machine shut-gown and plasma energy quench during a plasma disruption were carried out using 3D FEM code (ANSYS 5.2). (author). 5 tabs., 22 figs., 9 refs

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.

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

International Nuclear Information System (INIS)

Sanchez S, R.A.

2004-01-01

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

Natural science textbooks for the fourth grade and their text difficulty

Directory of Open Access Journals (Sweden)

Libuše Hrabí

2012-09-01

Full Text Available This paper presents findings regarding an assessment of the difficulty of text in six current Czech natural science textbooks for the fourth grade. The textual analysis was carried out according to a modified Průcha method. The results indicate that textual difficulty varies in the textbooks examined (19 - 31 points. Textbooks published by the Alter, Fortuna and SPN publishing companies are suitable for teaching in the fourth grade.

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.

Consequence evaluation of hypothetical reactor pressure vessel support failure

International Nuclear Information System (INIS)

Lu, S.C.; Holman, G.S.; Lambert, H.E.

1991-01-01

This paper describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. The structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports and that the SG supports and the RCP supports have sufficient design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas for further investigation and concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns. (author)

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

Consumerism and the Sister Carrie's American Dream%Consumerism and the Sister Carrie''s American Dream

Institute of Scientific and Technical Information of China (English)

卢亚丽

2017-01-01

From the aspect of consumerism to this text analyze Sister Carrie's"American dream"destruction. The author wholly and deeply analyzes the embodiment of consumerism in Dreiser's Sister Carrie and Dreiser's outlook and values under the effect of consumerism. To prove that the reason for destruction of Carrie's American dream is consumerism.

Vessel core seismic interaction for a fast reactor

International Nuclear Information System (INIS)

Martelli, A.; Maresca, G.

1984-01-01

This report deals with the analysis carried out in collaboration between ENEA and NIRA for optimizing the iterative procedure applied for the evaluation of the effects of the vessel core dynamic interaction for a fast reactor in the case of a earthquake. In fact, as shown in a previous report the convergence of such procedure was very slow for the design solution adopted for the PEC reactor, i.e. with a core restraint plate located close to the top of the core elements. This study, although performed making use of preliminary data (the same of the cited previous report) demonstrates that the convergence is fast if a suitable linear core model is applied in the first iteration linear calculations carried out by NIRA, with an intermediate stiffness with respect to those corresponding to the two limit models previously assumed and increased damping coefficients. Thus, the optimized iterative procedures is now applied in the PEC reactor block seismic verification analysis

IWR-solution for the ITER vacuum vessel assembly

Energy Technology Data Exchange (ETDEWEB)

Wu, H., E-mail: huapeng@lut.fi [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Handroos, H. [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Pela, P. [Tekes (Finland); Wang, Y. [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland)

2011-10-15

The assembly of ITER vacuum vessel (VV) is still a very big challenge as the process can only be done from inside the VV. The welding of the VV assembly is carried out using the dedicated robotic systems. The main functions of the robots are: (i) measuring the actual space between every two sectors, (ii) positioning of the 150 kg splice plates between the sector shells, (iii) welding the splice plates to the sector shells, (iv) NDT of the welds, (v) repairing, including machining of the welds, (vi) He-leak tests of the welds, and (vii) the non-planned functions that may turn out. This paper presents a reasonable method to assemble the ITER VV. In this article, one parallel mobile robot, running on the track rail fixed on the wall inside the VV, is designed and tested. The assembling process, carried out by the mobile robot together with the welding robot, is presented.

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

Expression of nm23-H1 gene product in esophageal squamous cell carcinoma and its association with vessel invasion and survival

International Nuclear Information System (INIS)

Tomita, Masaki; Ayabe, Takanori; Matsuzaki, Yasunori; Edagawa, Masao; Maeda, Masayuki; Shimizu, Tetsuya; Hara, Masaki; Onitsuka, Toshio

2001-01-01

We assessed the nm23-H1 gene product expression and its relationship with lymphatic and blood vessel invasion in patients with esophageal squamous cell carcinoma. Formalin-fixed and paraffin-embedded tissue sections from 45 patients who were treated surgically were used in this study. Pathologists graded lymphatic and blood vessel invasion in each of the tissue samples. Expression of nm23-Hl gene product was determined using a specific monoclonal antibody. Expression of nm23-H1 gene product was present in 17 (37.8%) cases. We found an inverse correlation between nm23-H1 gene product expression and lymphatic vessel invasion, whereas no correlation between nm23-H1 gene product expression and blood vessel invasion. Overall survival rate was not different between nm23-H1 gene product positive and negative patients (p = 0.21). However, reduced expression of nm23-H1 gene product was associated with shorter overall survival in patients with involved lymph nodes (p < 0.05), but not in patients without involved lymph nodes (p = 0.87). In patients with esophageal squamous cell carcinoma, there appears to be an inverse relationship between nm23-H1 gene product expression and lymphatic vessel invasion. Furthermore, nm23-H1 gene product expression might be a prognostic marker in patients with involved lymph nodes. Our data does not demonstrate any correlation between nm23-H1 gene product expression and blood vessel invasion

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)

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

Development of a reconstitution system of Charpy probes for the surveillance of vessels in nucleo electric plants

International Nuclear Information System (INIS)

Romero C, J.; Hernandez, R.; Fernandez, F.; Gonzalez M, A.

2007-01-01

This work describes the development of a welding system, for the rebuilding of halves of Charpy test tubes, the rebuilding consists on welding two implants in those ends of these halves of test tubes, in these welding the main requirement is not to alter the mechanical properties in a minimum volume of 1 cm 3 , the rebuilding is medullary in the surveillance programs of the reactor vessel. In these programs, the mechanical state of the vessel is evaluated, for it there are surveillance capsules with a Charpy witness test tubes series, subjected to a neutron flow similar or bigger to that of the vessel. The objective is to evaluate in advance on the vessel fragilization grade its life design. However the number of capsules with the witness test tubes it is only for the plant design life and at the moment the nucleo electric, negotiates an extension of life of these, until for 20 more years, of there the importance of this material witness's that stores the information of the damage accumulated by the neutron flow. This material requires to be taken advantage it after being rehearsed and the normative one settles down as obligatory to qualify the rebuilding process with all the requirements settled down in the ASTM Designation: E 1253-99 'Standard Guide for Reconstitution of irradiated Charpy-Sized Specimens', to obtain other reconstituted Charpy test tubes that are again introduced in the reactor. When being reconstituted the halves of the original test tubes it is obtained double reconstituted Charpy test tubes. Half of the test tubes they are used in the surveillance program of the vessel, with the surpluses test tubes, it can determine the fracture toughness, property of the material used in the extension methodology of life of vessel. (Author)

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

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

International Nuclear Information System (INIS)

Ng, D.S.

1986-09-01

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

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

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.)

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

Research Vessel R/V Sikuliaq: A New Asset For The UNOLS Fleet

Science.gov (United States)

Whitledge, T. E.

2012-12-01

The research vessel R/V Sikuliaq is currently being constructed on behalf of the NSF to support future scientific studies in high latitude waters. The 261 foot global class vessel will be capable of breaking 2.5 foot thick ice at 2 knots with an endurance of 45 days at sea and cruising at 11 knots. The R/V Sikuliaq will have a beam of 52 feet and a draft of 18.9 feet that will carry 26 scientists and a crew of 20. Berthing accommodations are a combination of single/double rooms with one stateroom and the common areas of the vessel are designed for ADA access and accommodations. The total laboratory space (main, analytical, electronics, wet, upper, and Baltic room will be 2100 square feet. The 4360 square foot working deck that is approximately 70 feet in length will accommodate 2-4 vans and multiple science operations. The vessel design strives to have the lowest possible environmental impact, including a low underwater-radiated noise signature. The science systems are prescribed to be state-of-the-art for bottom mapping, over-the-side "hands free" gear handling, broad band communications and scientific walk-in freezer and environmental chamber. More details and photos of the construction progress are available on the website at www.sfos.uaf.edu/arrv. The shipyard schedule has a launch date of October 2012 and delivery to the University of Alaska Fairbanks in July 2013. Scientific operations following trials and testing is planned to start in January 2014. Questions about the science systems or vessel capabilities should be directed to Terry Whitledge (terry@ims.uaf.edu).;

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

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

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)

Identification and assessment of BWR in-vessel severe accident mitigation strategies

Energy Technology Data Exchange (ETDEWEB)

Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. [Oak Ridge National Lab., TN (United States)

1992-10-01

This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

Identification and assessment of BWR in-vessel severe accident mitigation strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M.

1992-10-01

This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

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)

Approaches for accounting and prediction of fast neutron fluence on WWER pressure vessels and results of validation of calculational procedure

International Nuclear Information System (INIS)

Borodkin, P.G.; Khrennikov, N.N.; Ryabinin, Yu.A.; Adeev, V.A.

2015-01-01

A description is given of the universal procedure for calculation of fast neutron fluence (FNF) on WWER vessels. Approbation of the calculation procedure was carried out by comparing the calculation results for this procedure and measurements on the outer surface of the WWER-440 and WWER-1000 vessels. In addition, an estimation of the uncertainty of the settlement procedure was made in accordance with the requirements of regulatory documents. The developed procedure is applied at Kola NPP for independent fast neutron fluence estimates on the WWER-440 reactor vessels when planning core loads taking into account the introduction of new fuels. The results of the pilot operation of the procedure for calculating FNF at the Kola NPP were taken into account when improving the procedure and its application to the calculations of FNF on the WWER-1000 vessels [ru

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.)

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. �

In-vivo segmentation and quantification of coronary lesions by optical coherence tomography images for a lesion type definition and stenosis grading.

Science.gov (United States)

Celi, Simona; Berti, Sergio

2014-10-01

Optical coherence tomography (OCT) is a catheter-based medical imaging technique that produces cross-sectional images of blood vessels. This technique is particularly useful for studying coronary atherosclerosis. In this paper, we present a new framework that allows a segmentation and quantification of OCT images of coronary arteries to define the plaque type and stenosis grading. These analyses are usually carried out on-line on the OCT-workstation where measuring is mainly operator-dependent and mouse-based. The aim of this program is to simplify and improve the processing of OCT images for morphometric investigations and to present a fast procedure to obtain 3D geometrical models that can also be used for external purposes such as for finite element simulations. The main phases of our toolbox are the lumen segmentation and the identification of the main tissues in the artery wall. We validated the proposed method with identification and segmentation manually performed by expert OCT readers. The method was evaluated on ten datasets from clinical routine and the validation was performed on 210 images randomly extracted from the pullbacks. Our results show that automated segmentation of the vessel and of the tissue components are possible off-line with a precision that is comparable to manual segmentation for the tissue component and to the proprietary-OCT-console for the lumen segmentation. Several OCT sections have been processed to provide clinical outcome. Copyright © 2014 Elsevier B.V. All rights reserved.

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

International Nuclear Information System (INIS)

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

1985-08-01

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

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

Microstructure and properties of aluminium-aluminium oxide graded composite materials

Science.gov (United States)

Kamaruzaman, F. F.; Nuruzzaman, D. M.; Ismail, N. M.; Hamedon, Z.; Iqbal, A. K. M. A.; Azhari, A.

2018-03-01

In this research works, four-layered aluminium-aluminium oxide (Al-Al2O3) graded composite materials were fabricated using powder metallurgy (PM) method. In processing, metal-ceramic graded composite materials of 0%, 10%, 20% and 30% weight percentage of ceramic concentration were prepared under 30 ton compaction load using a cylindrical die-punch set made of steel. After that, two-step pressureless sintering was carried out at sintering temperature and time 600°C and 3 hours respectively. It was observed that the sintered cylindrical specimens of 30 mm diameter were prepared successfully. The graded composite specimens were analysed and the properties such as density, microstructure and hardness were measured. It was found that after sintering process, the diameter of the graded cylindrical structure was decreased. Using both Archimedes method and rule of mixture (ROM), he density of structure was measured. The obtained results revealed that the microvickers hardness was increased as the ceramic component increases in the graded layer. Moreover, it was observed that the interface of the graded structure is clearly distinguished within the multilayer stack and the ceramic particles are almost uniformly distributed in the Al matrix.

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.

The implementation of vessel-sinking policy as an effort to protect indonesian fishery resources and territorial waters

Science.gov (United States)

Nurdin; Ikaningtyas; Kurniaty, Rika

2018-04-01

This study aims to analysis the effectiveness of foreign ship sinking policies to eradicate illegal, unreported, and unregulated (IUU) fishing. There are many foreign fishing vessels were detained due to IUU fishing in Indonesia`s exclusive economic zone (EEZ) waters, particularly in the Natuna and Anambas region. In combating illegal fishing, the government of the Republic of Indonesia take concrete actions in protecting marine potentials by sinking foreign vessel policies. In the last three years more than 300 foreign ships are drowned by Indonesian government. This study revealed that regulations concerning the act of sinking the vessel have been in existence since 2009 but lack of socialization. The Indonesian government’s policy regarding foreign-flagged vessel carrying out IUU fishing is regulated under Law Number 45 of 2009 on Fisheries, and internationally permitted with certain restrictions on conditions set forth in article 73 paragraph (3) of UNCLOS 1982. These policy is part of an effort to improve the deterrence effect of regional offenses that could harm and threaten the sovereignty of the state.

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

An Experimental Study to Replace the Thoracic Descending Aorta for Pigs with a Self-Made Sutureless Blood Vessel

Directory of Open Access Journals (Sweden)

Fenglin Song

2014-01-01

Full Text Available To simplify the procedure of blood vessel replacement operation and shorten the vascular anastomosis time, we developed a special artificial blood vessel which can be connected to native blood vessels without suture. The self-made sutureless blood vessel (SMSBV was made from two titanium connectors and a Gore-Tex graft. To investigate blood compatibility and histocompatibility of the SMSBV, we carried thoracic descending aorta replacement using either SMSBV or Gore-Tex, respectively, in pigs. The aortic clamp time and the operative blood loss in the experimental group (using SMSBV were less than those in the control group (using Gore-Tex. The whole blood hematocrit, platelet count, plasma soluble P-selectin, plasma free hemoglobin, and interleukins 2, 6 at each time point were not different significantly between the two groups. Light microscopy and transmission electron microscopy examination showed there were layers of vascular smooth muscle cells and endothelial cells adhered in the inner wall of artificial blood vessel without any signs of thrombosis. Based on the result, we have drawn the conclusion that the application of SMSBV can significantly shorten the vascular anastomosis time, reduce operative blood loss, and show good blood and tissue compatibility.

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.)

Preventing the embrittling by hydrogen when galvanizing high-grade steel

Energy Technology Data Exchange (ETDEWEB)

Paatsch, W.

1987-09-01

Galvanic precipitation of a double layer consisting of a dull nickel layer overlaid with a brilliant zinc layer on low-alloyed high-strength steel grades leads to the forming of zinc-nickel alloy layers during the subsequent heat treatment. According to traction tests carried out on high-strength steel grades, as well as to hydrogen permeability tests, this process prevents embrittling by hydrogen which might be caused by galvanic process sequences - and creates a diffusion block at the same time. The alloy layers have an excellent corrosion resistance and temperature stability.

Vessels in Transit - Web Tool

Data.gov (United States)

Department of Transportation — A web tool that provides real-time information on vessels transiting the Saint Lawrence Seaway. Visitors may sort by order of turn, vessel name, or last location in...

High-performance fiber/epoxy composite pressure vessels

Science.gov (United States)

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

1978-01-01

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

Reactor vessel supported by flexure member

International Nuclear Information System (INIS)

Crawford, J.D.; Pankow, B.

1975-01-01

A description is given of a reactor pressure vessel which is provided with vertical support means in the form of circumferentially spaced columns upon which the vessel is mounted. The columns are adapted to undergo flexure in order to accommodate the thermally induced displacements experienced by the vessel during operational transients

The pattern of renal vessels in live related potential donors pool. A multislice computed tomography angiography review

International Nuclear Information System (INIS)

Mishra, A.; Ehtuish, Ehtuish F.

2006-01-01

To assess the renal vessel anatomy, compare the findings with the perioperative findings, to determine the sensitivity of multislice computed tomography (CT) angiography in the work-up of live potential donors and to discuss and compare the results of the present study with the reported results using single slice CT, magnetic resonance (MRI) and conventional angiography (CA).Retrospective analysis of the angiographic data of 118 of prospective live related kidney donors was carried out from October 2004 to August 2005 at the National Organ Transplant Centre, Tripoli Central Hospital, Libya. All donors underwent renal angiography on multislice (16-slice) CT scan using 80 cc intravenous contrast with 1.25 mm slice thickness followed by maximum intensity projection (MIP) and volume rendering techniques (VRT) post-processing algorithms. The number of vessels, vessel bifurcation, vessel morphology and venous anatomy were analyzed and the findings were compared with the surgical findings. Multislice spiral CT angiography (MSCTA) showed clear delineation of the main renal arteries in all donors with detailed vessel morphology. The study revealed 100% sensitivity in detection of accessory renal vessels, with an overall incidence of 26.7%, which is the most common distribution in the parahilar region. The present study showed 100% sensitivity in the visualization and detection of main and accessory renal vessels. These results were comparable with conventional angiography which has so far been considered as the gold standard and were found superior in specificity and accuracy to the use of single slice CT (SSCT) and MR in the angiographic work-up of live renal donors. Due to improved detection of accessory vessels less than 2 mm in diameter, a higher incidence of aberrant vessels was seen on the right side as has been suggested so far. (author)

Low cycle fatigue behavior of a ferritic reactor pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Sarkar, Apu, E-mail: asarkar@barc.gov.in; Kumawat, Bhupendra K.; Chakravartty, J.K.

2015-07-15

The cyclic stress–strain response and the low cycle fatigue (LCF) behavior of 20MnMoNi55 pressure vessel steel were studied. Tensile strength and LCF properties were examined at room temperature (RT) using specimens cut from rolling direction of a rolled block. The fully reversed strain-controlled LCF tests were conducted at a constant total strain rate with different axial strain amplitude levels. The cyclic strain–stress relationships and the strain–life relationships were obtained through the test results, and related LCF parameters of the steel were calculated. The studied steel exhibits cyclic softening behavior. Furthermore, analysis of stabilized hysteresis loops showed that the steel exhibits non-Masing behavior. Complementary scanning electron microscopy examinations were also carried out on fracture surfaces to reveal dominant damage mechanisms during crack initiation, propagation and fracture. Multiple crack initiation sites were observed on the fracture surface. The investigated LCF behavior can provide reference for pressure vessel life assessment and fracture mechanisms analysis.

Relationship between Legible Handwriting and Level of Success of Third Grade Students in Written Expression

Science.gov (United States)

Bayat, Seher; Küçükayar, Hasan

2016-01-01

This study aims to identify third-grade students' performance levels for written expression and handwriting and to find the relationship between these performances. The study is based on relational screening model. It is carried out with 110 third grade students. Students' levels of success in handwriting and in written expression are evaluated…

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Comprehensive, Integrative Genomic Analysis of Diffuse Lower-Grade Gliomas.

Science.gov (United States)

Brat, Daniel J; Verhaak, Roel G W; Aldape, Kenneth D; Yung, W K Alfred; Salama, Sofie R; Cooper, Lee A D; Rheinbay, Esther; Miller, C Ryan; Vitucci, Mark; Morozova, Olena; Robertson, A Gordon; Noushmehr, Houtan; Laird, Peter W; Cherniack, Andrew D; Akbani, Rehan; Huse, Jason T; Ciriello, Giovanni; Poisson, Laila M; Barnholtz-Sloan, Jill S; Berger, Mitchel S; Brennan, Cameron; Colen, Rivka R; Colman, Howard; Flanders, Adam E; Giannini, Caterina; Grifford, Mia; Iavarone, Antonio; Jain, Rajan; Joseph, Isaac; Kim, Jaegil; Kasaian, Katayoon; Mikkelsen, Tom; Murray, Bradley A; O'Neill, Brian Patrick; Pachter, Lior; Parsons, Donald W; Sougnez, Carrie; Sulman, Erik P; Vandenberg, Scott R; Van Meir, Erwin G; von Deimling, Andreas; Zhang, Hailei; Crain, Daniel; Lau, Kevin; Mallery, David; Morris, Scott; Paulauskis, Joseph; Penny, Robert; Shelton, Troy; Sherman, Mark; Yena, Peggy; Black, Aaron; Bowen, Jay; Dicostanzo, Katie; Gastier-Foster, Julie; Leraas, Kristen M; Lichtenberg, Tara M; Pierson, Christopher R; Ramirez, Nilsa C; Taylor, Cynthia; Weaver, Stephanie; Wise, Lisa; Zmuda, Erik; Davidsen, Tanja; Demchok, John A; Eley, Greg; Ferguson, Martin L; Hutter, Carolyn M; Mills Shaw, Kenna R; Ozenberger, Bradley A; Sheth, Margi; Sofia, Heidi J; Tarnuzzer, Roy; Wang, Zhining; Yang, Liming; Zenklusen, Jean Claude; Ayala, Brenda; Baboud, Julien; Chudamani, Sudha; Jensen, Mark A; Liu, Jia; Pihl, Todd; Raman, Rohini; Wan, Yunhu; Wu, Ye; Ally, Adrian; Auman, J Todd; Balasundaram, Miruna; Balu, Saianand; Baylin, Stephen B; Beroukhim, Rameen; Bootwalla, Moiz S; Bowlby, Reanne; Bristow, Christopher A; Brooks, Denise; Butterfield, Yaron; Carlsen, Rebecca; Carter, Scott; Chin, Lynda; Chu, Andy; Chuah, Eric; Cibulskis, Kristian; Clarke, Amanda; Coetzee, Simon G; Dhalla, Noreen; Fennell, Tim; Fisher, Sheila; Gabriel, Stacey; Getz, Gad; Gibbs, Richard; Guin, Ranabir; Hadjipanayis, Angela; Hayes, D Neil; Hinoue, Toshinori; Hoadley, Katherine; Holt, Robert A; Hoyle, Alan P; Jefferys, Stuart R; Jones, Steven; Jones, Corbin D; Kucherlapati, Raju; Lai, Phillip H; Lander, Eric; Lee, Semin; Lichtenstein, Lee; Ma, Yussanne; Maglinte, Dennis T; Mahadeshwar, Harshad S; Marra, Marco A; Mayo, Michael; Meng, Shaowu; Meyerson, Matthew L; Mieczkowski, Piotr A; Moore, Richard A; Mose, Lisle E; Mungall, Andrew J; Pantazi, Angeliki; Parfenov, Michael; Park, Peter J; Parker, Joel S; Perou, Charles M; Protopopov, Alexei; Ren, Xiaojia; Roach, Jeffrey; Sabedot, Thaís S; Schein, Jacqueline; Schumacher, Steven E; Seidman, Jonathan G; Seth, Sahil; Shen, Hui; Simons, Janae V; Sipahimalani, Payal; Soloway, Matthew G; Song, Xingzhi; Sun, Huandong; Tabak, Barbara; Tam, Angela; Tan, Donghui; Tang, Jiabin; Thiessen, Nina; Triche, Timothy; Van Den Berg, David J; Veluvolu, Umadevi; Waring, Scot; Weisenberger, Daniel J; Wilkerson, Matthew D; Wong, Tina; Wu, Junyuan; Xi, Liu; Xu, Andrew W; Yang, Lixing; Zack, Travis I; Zhang, Jianhua; Aksoy, B Arman; Arachchi, Harindra; Benz, Chris; Bernard, Brady; Carlin, Daniel; Cho, Juok; DiCara, Daniel; Frazer, Scott; Fuller, Gregory N; Gao, JianJiong; Gehlenborg, Nils; Haussler, David; Heiman, David I; Iype, Lisa; Jacobsen, Anders; Ju, Zhenlin; Katzman, Sol; Kim, Hoon; Knijnenburg, Theo; Kreisberg, Richard Bailey; Lawrence, Michael S; Lee, William; Leinonen, Kalle; Lin, Pei; Ling, Shiyun; Liu, Wenbin; Liu, Yingchun; Liu, Yuexin; Lu, Yiling; Mills, Gordon; Ng, Sam; Noble, Michael S; Paull, Evan; Rao, Arvind; Reynolds, Sheila; Saksena, Gordon; Sanborn, Zack; Sander, Chris; Schultz, Nikolaus; Senbabaoglu, Yasin; Shen, Ronglai; Shmulevich, Ilya; Sinha, Rileen; Stuart, Josh; Sumer, S Onur; Sun, Yichao; Tasman, Natalie; Taylor, Barry S; Voet, Doug; Weinhold, Nils; Weinstein, John N; Yang, Da; Yoshihara, Kosuke; Zheng, Siyuan; Zhang, Wei; Zou, Lihua; Abel, Ty; Sadeghi, Sara; Cohen, Mark L; Eschbacher, Jenny; Hattab, Eyas M; Raghunathan, Aditya; Schniederjan, Matthew J; Aziz, Dina; Barnett, Gene; Barrett, Wendi; Bigner, Darell D; Boice, Lori; Brewer, Cathy; Calatozzolo, Chiara; Campos, Benito; Carlotti, Carlos Gilberto; Chan, Timothy A; Cuppini, Lucia; Curley, Erin; Cuzzubbo, Stefania; Devine, Karen; DiMeco, Francesco; Duell, Rebecca; Elder, J Bradley; Fehrenbach, Ashley; Finocchiaro, Gaetano; Friedman, William; Fulop, Jordonna; Gardner, Johanna; Hermes, Beth; Herold-Mende, Christel; Jungk, Christine; Kendler, Ady; Lehman, Norman L; Lipp, Eric; Liu, Ouida; Mandt, Randy; McGraw, Mary; Mclendon, Roger; McPherson, Christopher; Neder, Luciano; Nguyen, Phuong; Noss, Ardene; Nunziata, Raffaele; Ostrom, Quinn T; Palmer, Cheryl; Perin, Alessandro; Pollo, Bianca; Potapov, Alexander; Potapova, Olga; Rathmell, W Kimryn; Rotin, Daniil; Scarpace, Lisa; Schilero, Cathy; Senecal, Kelly; Shimmel, Kristen; Shurkhay, Vsevolod; Sifri, Suzanne; Singh, Rosy; Sloan, Andrew E; Smolenski, Kathy; Staugaitis, Susan M; Steele, Ruth; Thorne, Leigh; Tirapelli, Daniela P C; Unterberg, Andreas; Vallurupalli, Mahitha; Wang, Yun; Warnick, Ronald; Williams, Felicia; Wolinsky, Yingli; Bell, Sue; Rosenberg, Mara; Stewart, Chip; Huang, Franklin; Grimsby, Jonna L; Radenbaugh, Amie J; Zhang, Jianan

2015-06-25

Diffuse low-grade and intermediate-grade gliomas (which together make up the lower-grade gliomas, World Health Organization grades II and III) have highly variable clinical behavior that is not adequately predicted on the basis of histologic class. Some are indolent; others quickly progress to glioblastoma. The uncertainty is compounded by interobserver variability in histologic diagnosis. Mutations in IDH, TP53, and ATRX and codeletion of chromosome arms 1p and 19q (1p/19q codeletion) have been implicated as clinically relevant markers of lower-grade gliomas. We performed genomewide analyses of 293 lower-grade gliomas from adults, incorporating exome sequence, DNA copy number, DNA methylation, messenger RNA expression, microRNA expression, and targeted protein expression. These data were integrated and tested for correlation with clinical outcomes. Unsupervised clustering of mutations and data from RNA, DNA-copy-number, and DNA-methylation platforms uncovered concordant classification of three robust, nonoverlapping, prognostically significant subtypes of lower-grade glioma that were captured more accurately by IDH, 1p/19q, and TP53 status than by histologic class. Patients who had lower-grade gliomas with an IDH mutation and 1p/19q codeletion had the most favorable clinical outcomes. Their gliomas harbored mutations in CIC, FUBP1, NOTCH1, and the TERT promoter. Nearly all lower-grade gliomas with IDH mutations and no 1p/19q codeletion had mutations in TP53 (94%) and ATRX inactivation (86%). The large majority of lower-grade gliomas without an IDH mutation had genomic aberrations and clinical behavior strikingly similar to those found in primary glioblastoma. The integration of genomewide data from multiple platforms delineated three molecular classes of lower-grade gliomas that were more concordant with IDH, 1p/19q, and TP53 status than with histologic class. Lower-grade gliomas with an IDH mutation either had 1p/19q codeletion or carried a TP53 mutation. Most

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

Energy Technology Data Exchange (ETDEWEB)

Heel, A.M.J.M. van

1995-07-01

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

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

International Nuclear Information System (INIS)

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

1995-07-01

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

Planary bone scintigraphy of the foot in the diabetics with peripheral macroangiopathy, treated with Sulodexide (Vessel Due F)

International Nuclear Information System (INIS)

Klisarova, A.; Bihchelian, H.; Koeva, L.

2000-01-01

It is the purpose of the paper to assay the effect of Sulodexide (Vessel Due F) on some semi-quantitative parameters of bone scintigraphy of the foot in diabetics with peripheral vascular disease. Fifteen patients with diabet type II and peripheral macroangiopathy (6 women and 9 men, mean age 57 ±3.7 y, and mean body mass index 29.8 ±2.7 kg/m 3 ) are studied. The investigation is carried out after informed consent and in state of adequate glycemic control. Vessel Due F is administrated im for 10 days - 600 lypoproteinli pase releasing units (LSU), followed by 500 LSU per os for further 60 days. Before and after treatment, planar foot bone scintigraphy with 99m Tc - MDP is carried out. The fixation indices are measured through comparison of radionuclide accumulation in symmetrical zones of the right and left foot. Prior to treatment the fixation indices are increased, and after treatment they are significantly decreased (p< 0.05). The semiquantitative scintigraphic parameters adopted are used for dynamic measurement of the bone metabolism level in the feet of patients with diabet and peripheral macroangiopathy. After Sulodexide (Vessel Due F) treatment, a positive effect on radionuclide fixation indices is documented. (author)

19 CFR 4.97 - Salvage vessels.

Science.gov (United States)

2010-04-01

... United States and Great Britain ‘concerning reciprocal rights for United States and Canada in the... meaning of this statute. (e) A Mexican vessel may engage in a salvage operation on a Mexican vessel in any territorial waters of the United States in which Mexican vessels are permitted to conduct such operations by...

Vacuum vessel for thermonuclear device

International Nuclear Information System (INIS)

Hagiwara, Koji; Imura, Yasuya.

1979-01-01

Purpose: To provide constituted method for easily performing baking of vacuum vessel, using short-circuiting segments. Constitution: At the time of baking, one turn circuit is formed by the vacuum vessel and short-circuiting segments, and current transformer converting the one turn circuit into a secondary circuit by the primary coil and iron core is formed, and the vacuum vessel is Joule heated by an induction current from the primary coil. After completion of baking, the short-circuiting segments are removed. (Kamimura, M.)

Structural analysis of the KSTAR vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

In, Sang Ryul; Yoon, Byeong Joo [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-10-01

Structure analysis of the vacuum vessel for the KSTAR tokamak which, is in the end phase of the conceptual design have been performed. Mechanical stresses and deformations of the vessel produced by constant forces due to atmospheric pressure, dead weight, fluid pressure, etc and various transient electromagnetic forces induced during tokamak operations were calculated as well as modal characteristics and buckling properties were investigated. Influences of the temperature gradient and the constraint condition of the support on the thermal stress and deformation of the vessel were analyzed. The thermal stress due to the temperature distribution on the vessel as supplying the N{sub 2} gas of 400 deg C through poloidal channels according to the recent baking concept were calculated. No severe problem in the robustness of the vessel was found when applying the constant pressures on the vessel. However the mechanical stress due to the EM force induced by halo currents flowing on the vessel and the plasma facing components (PFCs) far exceeded the allowable limit. Some reinforcing components should be added on the boundary of the PFC support and the vessel, and that of the vessel support and the vessel. A steep temperature gradient in the vicinity of the inlet and oulet of the heating gas produced a thermal stress much higher than allowable. It is necessary to make the temperature of the vessel as uniform as possible and to develop a new support concept which is flexible enough to accommodate a thermal expansion of a few cm while sufficiently strong to resist mechanical impacts. (author). 5 refs., 41 figs., 9 tabs.

Splenic trauma management in relation to mode and grade

International Nuclear Information System (INIS)

Gangat, S.A.; Khaskhali, A.A.; Memon, I.A.

2008-01-01

To study the prevalence and management of splenic trauma in relation to its mode and grade. All cases admitted in emergency with abdominal trauma and splenic injury. The data of all the patients who had splenic trauma was entered on a proforma and analyzed. A total of 44 patients with ages between 20-40 years presented with splenic injury; 32(72.7%) were male. The commonest mode of splenic trauma was blunt abdominal injury (50%), and most (47%) patients had Grade- III injury. Splenectomy was carried out in 84% patients, while 9% underwent splenic salvage. Seven (15.9%) patients with splenectomy died in the series. Splenic injury was mostly caused by blunt abdominal trauma. Proper assessment of the grade of injury at the time of laparotomy resulted in more splenic salvage procedures with decreased risk of complications. (author)

Pressure vessels and methods of sealing leaky tubes disposed in pressure vessels

International Nuclear Information System (INIS)

Larson, G.C.

1980-01-01

This invention relates to pressure vessels and to methods of sealing leaky tubes in them and is especially applicable to pressure vessels in the form of sheet-and-tube type heat exchangers constructed with a large number of relatively small diameter tubes grouped in a bundle. To seal off a leaky tube in such a heat exchanger an explosive activated plug in the form of a hollow metal body is used, inserted at each end of the tube to be sealed. Using the arrangement of pressure vessel and associated tube sheets and the explosive activated plug method of sealing a leaky tube as described in this invention it is claimed that distortion of the adjacent tubes and the tube sheets is reduced when the explosive activated plugs are detonated. (U.K.)

Behavioural responses of dusky dolphin groups (Lagenorhynchus obscurus to tour vessels off Kaikoura, New Zealand.

Directory of Open Access Journals (Sweden)

David Lundquist

Full Text Available BACKGROUND: Commercial viewing and swimming with dusky dolphins (Lagenorhynchus obscurus near Kaikoura, New Zealand began in the late 1980s and researchers have previously described changes in vocalisation, aerial behaviour, and group spacing in the presence of vessels. This study was conducted to assess the current effects that tourism has on the activity budget of dusky dolphins to provide wildlife managers with information for current decision-making and facilitate development of quantitative criteria for management of this industry in the future. METHODOLOGY/PRINCIPAL FINDINGS: First-order time discrete Markov chain models were used to assess changes in the behavioural state of dusky dolphin pods targeted by tour vessels. Log-linear analysis was conducted on behavioural state transitions to determine whether the likelihood of dolphins moving from one behavioural state to another changed based on natural and anthropogenic factors. The best-fitting model determined by Akaike Information Criteria values included season, time of day, and vessel presence within 300 m. Interactions with vessels reduced the proportion of time dolphins spent resting in spring and summer and increased time spent milling in all seasons except autumn. Dolphins spent more time socialising in spring and summer, when conception occurs and calves are born, and the proportion of time spent resting was highest in summer. Resting decreased and traveling increased in the afternoon. CONCLUSIONS/SIGNIFICANCE: Responses to tour vessel traffic are similar to those described for dusky dolphins elsewhere. Disturbance linked to vessels may interrupt social interactions, carry energetic costs, or otherwise affect individual fitness. Research is needed to determine if increased milling is a result of acoustic masking of communication due to vessel noise, and to establish levels at which changes to behavioural budgets of dusky dolphins are likely to cause long-term harm. Threshold

33 CFR 151.1512 - Vessel safety.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Vessel safety. 151.1512 Section... River § 151.1512 Vessel safety. Nothing in this subpart relieves the master of the responsibility for ensuring the safety and stability of the vessel or the safety of the crew and passengers, or any other...

Deformation of cylindrical vessel and the effect of barrel on deformation under inpulsive pressure of high explosive

International Nuclear Information System (INIS)

Iikura, Shoichi; Yashizawa, Hiroyasu; Sasanuma, Katsumi.

1982-01-01

According to the research performed so far, the result that the amount of deformation due to impulsive pressure was able to be evaluated by the impulse of impulsive pressure waves has been obtained. The analysis treating impulsive pressure waves as plane waves has been made frequently, but the analysis in which impulsive pressure waves must be treated as spherical waves, or the analysis of a vessel with a barrel (internal cylinder) is complex and difficult. In this report, the results of element test, which was carried out in the Oita Works, Asahi Chemical Industry Co., Ltd., in 1973 by the Power Reactor and Nuclear Fuel Development Corp. as the impact resistance test for fast breeder reactors, are rearranged and investigated. The specimens were the cylindrical vessels with upper and lower flanges, and 10 vessels and 9 kinds of barrels were made. Water was used as the pressure medium. The residual deformation and dynamic strain of the vessels and the wave form of pressure waves were measured. The deformation of cylindrical vessels subjected to the impulsive pressure from a point pressure source was able to be evaluated by the impulse distribution in normal direction. The maximum amount of deformation depended on the total plate thickness of barrels. (Kako, I.)

Vitamin D Status in Small Vessel and Large Vessel Ischemic Stroke Patients: A Case–control Study

Directory of Open Access Journals (Sweden)

Navid Manouchehri

2017-01-01

Full Text Available Background: Vitamin D insufficiency is a globally widespread issue. Recent studies have reported a high prevalence of Vitamin D deficiency in Middle-East countries. Studies have shown negative effects of Vitamin D deficiency on endothelium and related diseases such as ischemic brain stroke. Here, we assessed Vitamin D status in patients with different types of ischemic brain stroke and control group. Materials and Methods: Seventy-five patients (49.3% small vessel, 50.7% large vessel and 75 controls, matched for age (68.01 ± 10.94 vs. 67.64 ± 10.24 and sex (42 male and 33 female were recruited. 25(OH D levels were measured by Chemiluminescence immunoassay. 25(OH D status was considered as severely, moderately, or mildly deficient and normal with 25(OH D levels of less than 5, 5-10, 10-16, and> 16 ng/ml, respectively. Results: Mean ± standard error concentration of 25(OH D in cases and controls were 17.7 ± 1.5 and 26.9 ± 1.6 (P = 0.0001, respectively. Mild, moderate, and severe Vitamin D deficiency were observed in 10.8%, 32.4%, 8.1% vs. 34.3%, 31.5%, 9.5% of small vessel and large vessel group, respectively. 21.7% of the controls were Vitamin D deficient. Vitamin D deficiency was significantly associated with higher risk for ischemic stroke, (P = 0.000, OR = 7.17, 95% confidence interval: 3.36–15.29. 25(OH D levels were significantly higher in control group comparing to small vessel (26.9 ± 1.6 vs. 20.59 ± 2.6 P < 0.05 and large vessel (26.9 ± 1.6 vs. 13.4 ± 1.3 P < 0.001 stroke patients. Small vessel group had significantly higher levels of Vitamin D than large vessel (P < 0.05. Conclusion: Vitamin D deficiency significantly increases the risk of ischemic stroke, favoring the types with the pathogenesis of large vessel strokes.

Feasibility Studies for Production of Pellet Grade Concentrate from Sub Grade Iron Ore Using Multi Gravity Separator

Science.gov (United States)

Rao, Gottumukkala Venkateswara; Markandeya, R.; Kumar, Rajan

2018-04-01

An attempt has been made to utilise Sub Grade Iron Ore by producing pellet grade concentrate from Deposit 5, Bacheli Complex, Bailadila, Chhattisgarh, India. The `as received' Run of Mine (ROM) sample assayed 40.80% Fe, 40.90% SiO2. Mineralogical studies indicated that the main ore mineral is Hematite and lone gangue mineral is Quartz. Mineral liberation studies indicated that, the ore mineral Hematite and gangue mineral Quartz are getting liberated below 100 microns. The stage crushed and ground sample was subjected to concentration by using a Multi Gravity Separator (MGS). Rougher Multi Gravity Separation (MGS) experimental results were optimised to recover highest possible iron values. A concentrate of 55.80% Fe with a yield of 61.73% by weight with a recovery of 84.42% Iron values was obtained in rougher MGS concentrate. Further experiments were carried out with rougher MGS concentrate to produce a concentrate suitable for commercial grade pellet concentrate. It was proved that a concentrate assaying 66.67% Fe, 3.12% SiO2 with an yield of 45.08% by weight and with a recovery of 73.67% iron values in the concentrate.

Effect of a new specimen size on fatigue crack growth behavior in thick-walled pressure vessels

International Nuclear Information System (INIS)

Shariati, Mahmoud; Mohammadi, Ehsan; Masoudi Nejad, Reza

2017-01-01

Fatigue crack growth in thick-walled pressure vessels is an important factor affecting their fracture. Predicting the path of fatigue crack growth in a pressure vessel is the main issue discussed in fracture mechanics. The objective of this paper is to design a new geometrical specimen in fatigue to define the behavior of semi-elliptical crack growth in thick-walled pressure vessels. In the present work, the importance of the behavior of fatigue crack in test specimen and real conditions in thick-walled pressure vessels is investigated. The results of fatigue loading on the new specimen are compared with the results of fatigue loading in a cylindrical pressure vessel and a standard specimen. Numerical and experimental methods are used to investigate the behavior of fatigue crack growth in the new specimen. For this purpose, a three-dimensional boundary element method is used for fatigue crack growth under stress field. The modified Paris model is used to estimate fatigue crack growth rates. In order to verify the numerical results, fatigue test is carried out on a couple of specimens with a new geometry made of ck45. A comparison between experimental and numerical results has shown good agreement. - Highlights: • This paper provides a new specimen to define the behavior of fatigue crack growth. • We estimate the behavior of fatigue crack growth in specimen and pressure vessel. • A 3D finite element model has been applied to estimate the fatigue life. • We compare the results of fatigue loading for cylindrical vessel and specimens. • Comparison between experimental and numerical results has shown a good agreement.

33 CFR 161.4 - Requirement to carry the rules.

Science.gov (United States)

2010-07-01

... (CONTINUED) PORTS AND WATERWAYS SAFETY VESSEL TRAFFIC MANAGEMENT Vessel Traffic Services General Rules § 161... the Local Notice to Mariners. The VTS User's Manual and the World VTS Guide, an International Maritime...

AFSC/FMA/Vessel Assessment Logging

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Vessels fishing trawl gear, vessels fishing hook-and-line and pot gear that are also greater than 57.5 feet overall, and shoreside and floating processing facilities...

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)

Assessment of environmentally assisted cracking in PWR pressure vessel steels

International Nuclear Information System (INIS)

Tice, D.R.

1991-01-01

There is a possibility that extension of pre-existing flaws in the reactor pressure vessel of a pressurised water reactor (PWR) may occur by environmentally assisted cracking, in particular by corrosion fatigue under cyclic transient loading. Crack growth predictions have usually been carried out using cyclic crack growth rate (da/dN) versus stress intensity range (δK) curves, such as those given in Section XI, Appendix A of the ASME Boiler and Pressure Vessel Code. However, the inherent time dependent nature of environmental cracking processes renders such an approach unrealistic. The present paper describes the development of an alternative time based assessment methodology. Illustrative calculations of expected crack growth of assumed defects made using the cyclic (ASME XIA) and time-based approaches are compared. The results illustrate that crack growth predicted by the time-based approach can be greater or less than that calculated by the traditional method. For a PWR operated with good control of water chemistry, actual crack growth rates are expected to be well below those predicted by the ASME code. (Author)

Structural and functional features of central nervous system lymphatic vessels.

Science.gov (United States)

Louveau, Antoine; Smirnov, Igor; Keyes, Timothy J; Eccles, Jacob D; Rouhani, Sherin J; Peske, J David; Derecki, Noel C; Castle, David; Mandell, James W; Lee, Kevin S; Harris, Tajie H; Kipnis, Jonathan

2015-07-16

One of the characteristics of the central nervous system is the lack of a classical lymphatic drainage system. Although it is now accepted that the central nervous system undergoes constant immune surveillance that takes place within the meningeal compartment, the mechanisms governing the entrance and exit of immune cells from the central nervous system remain poorly understood. In searching for T-cell gateways into and out of the meninges, we discovered functional lymphatic vessels lining the dural sinuses. These structures express all of the molecular hallmarks of lymphatic endothelial cells, are able to carry both fluid and immune cells from the cerebrospinal fluid, and are connected to the deep cervical lymph nodes. The unique location of these vessels may have impeded their discovery to date, thereby contributing to the long-held concept of the absence of lymphatic vasculature in the central nervous system. The discovery of the central nervous system lymphatic system may call for a reassessment of basic assumptions in neuroimmunology and sheds new light on the aetiology of neuroinflammatory and neurodegenerative diseases associated with immune system dysfunction.

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.

Vessel size measurements in angiograms: Manual measurements

International Nuclear Information System (INIS)

Hoffmann, Kenneth R.; Dmochowski, Jacek; Nazareth, Daryl P.; Miskolczi, Laszlo; Nemes, Balazs; Gopal, Anant; Wang Zhou; Rudin, Stephen; Bednarek, Daniel R.

2003-01-01

Vessel size measurement is perhaps the most often performed quantitative analysis in diagnostic and interventional angiography. Although automated vessel sizing techniques are generally considered to have good accuracy and precision, we have observed that clinicians rarely use these techniques in standard clinical practice, choosing to indicate the edges of vessels and catheters to determine sizes and calibrate magnifications, i.e., manual measurements. Thus, we undertook an investigation of the accuracy and precision of vessel sizes calculated from manually indicated edges of vessels. Manual measurements were performed by three neuroradiologists and three physicists. Vessel sizes ranged from 0.1-3.0 mm in simulation studies and 0.3-6.4 mm in phantom studies. Simulation resolution functions had full-widths-at-half-maximum (FWHM) ranging from 0.0 to 0.5 mm. Phantom studies were performed with 4.5 in., 6 in., 9 in., and 12 in. image intensifier modes, magnification factor = 1, with and without zooming. The accuracy and reproducibility of the measurements ranged from 0.1 to 0.2 mm, depending on vessel size, resolution, and pixel size, and zoom. These results indicate that manual measurements may have accuracies comparable to automated techniques for vessels with sizes greater than 1 mm, but that automated techniques which take into account the resolution function should be used for vessels with sizes smaller than 1 mm

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

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.

Performance of growing Yankasa rams Fed graded levels of ...

African Journals Online (AJOL)

A feeding trial which lasted eight (8) weeks was carried out to determine the intake and nutrient digestibility by growing Yankasa rams fed graded levels of Tamarindus indica leaves. Twelve Yankasa rams with average liveweight of 17.40kg were randomly allocated to three treatments of four replicates in a Randomized ...

Neutron Assay System for Confinement Vessel Disposition

International Nuclear Information System (INIS)

Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Valdez, Jose I.; Vigil, Georgiana M.

2012-01-01

Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the CVs. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of special nuclear material (SNM) in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of (le)100-g 239 Pu equivalent in a vessel for safeguards termination. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements.

Containment vessel

International Nuclear Information System (INIS)

Zbirohowski-Koscia, K.F.; Roberts, A.C.

1980-01-01

A concrete containment vessel for nuclear reactors is disclosed that is spherical and that has prestressing tendons disposed in first, second and third sets, the tendons of each set being all substantially concentric and centred around a respective one of the three orthogonal axes of the sphere; the tendons of the first set being anchored at each end at a first anchor rib running around a circumference of the vessel, the tendons of the second set being anchored at each end at a second anchor rib running around a circumference of the sphere and disposed at 90 0 to the first rib, and the tendons of the third set being anchored some to the first rib and the remainder to the second rib. (author)

7 CFR 810.2204 - Grades and grade requirements for wheat.

Science.gov (United States)

2010-01-01

... 7 Agriculture 7 2010-01-01 2010-01-01 false Grades and grade requirements for wheat. 810.2204... OFFICIAL UNITED STATES STANDARDS FOR GRAIN United States Standards for Wheat Principles Governing the Application of Standards § 810.2204 Grades and grade requirements for wheat. (a) Grades and grade requirements...

Stability, structure and scale: improvements in multi-modal vessel extraction for SEEG trajectory planning.

Science.gov (United States)

Zuluaga, Maria A; Rodionov, Roman; Nowell, Mark; Achhala, Sufyan; Zombori, Gergely; Mendelson, Alex F; Cardoso, M Jorge; Miserocchi, Anna; McEvoy, Andrew W; Duncan, John S; Ourselin, Sébastien

2015-08-01

Brain vessels are among the most critical landmarks that need to be assessed for mitigating surgical risks in stereo-electroencephalography (SEEG) implantation. Intracranial haemorrhage is the most common complication associated with implantation, carrying significantly associated morbidity. SEEG planning is done pre-operatively to identify avascular trajectories for the electrodes. In current practice, neurosurgeons have no assistance in the planning of electrode trajectories. There is great interest in developing computer-assisted planning systems that can optimise the safety profile of electrode trajectories, maximising the distance to critical structures. This paper presents a method that integrates the concepts of scale, neighbourhood structure and feature stability with the aim of improving robustness and accuracy of vessel extraction within a SEEG planning system. The developed method accounts for scale and vicinity of a voxel by formulating the problem within a multi-scale tensor voting framework. Feature stability is achieved through a similarity measure that evaluates the multi-modal consistency in vesselness responses. The proposed measurement allows the combination of multiple images modalities into a single image that is used within the planning system to visualise critical vessels. Twelve paired data sets from two image modalities available within the planning system were used for evaluation. The mean Dice similarity coefficient was 0.89 ± 0.04, representing a statistically significantly improvement when compared to a semi-automated single human rater, single-modality segmentation protocol used in clinical practice (0.80 ± 0.03). Multi-modal vessel extraction is superior to semi-automated single-modality segmentation, indicating the possibility of safer SEEG planning, with reduced patient morbidity.

Oxidation effect on steel corrosion and thermal loads during corium melt in-vessel retention

Energy Technology Data Exchange (ETDEWEB)

Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A.; Almjashev, V.I. [Alexandrov Scientific-Research Technology Institute (NITI), Sosnovy Bor (Russian Federation); Bechta, S.V. [KTH, Stockholm (Sweden); Gusarov, V.V. [SPb State Technology University (SPbGTU), St. Petersburg (Russian Federation); Barrachin, M. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), St Paul lez Durance (France); Bottomley, P.D., E-mail: paul.bottomley@ec.europa.eu [EC-Joint Research Centre, Institute for Transuranium Elements (ITU), Karlsruhe (Germany); Fischer, M. [AREVA GmbH, Erlangen (Germany); Piluso, P. [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), Cadarache, St Paul lez Durance (France)

2014-10-15

Highlights: • The METCOR facility simulates vessel steel corrosion in contact with corium. • Steel corrosion rates in UO{sub 2+x}–ZrO{sub 2}–FeO{sub y} coria accelerate above 1050 K. • However corrosion rates can also be limited by melt O{sub 2} supply. • The impact of this on in-vessel retention (IVR) strategy is discussed. - Abstract: During a severe accident with core meltdown, the in-vessel molten core retention is challenged by the vessel steel ablation due to thermal and physicochemical interaction of melt with steel. In accidents with oxidizing atmosphere above the melt surface, a low melting point UO{sub 2+x}–ZrO{sub 2}–FeO{sub y} corium pool can form. In this case ablation of the RPV steel interacting with the molten corium is a corrosion process. Experiments carried out within the International Scientific and Technology Center's (ISTC) METCOR Project have shown that the corrosion rate can vary and depends on both surface temperature of the RPV steel and oxygen potential of the melt. If the oxygen potential is low, the corrosion rate is controlled by the solid phase diffusion of Fe ions in the corrosion layer. At high oxygen potential and steel surface layer temperature of 1050 °C and higher, the corrosion rate intensifies because of corrosion layer liquefaction and liquid phase diffusion of Fe ions. The paper analyzes conditions under which corrosion intensification occurs and can impact on in-vessel melt retention (IVR)

Investigation of Horizontal Velocity Fields in Stirred Vessels with Helical Coils by PIV

Directory of Open Access Journals (Sweden)

Volker Bliem

2014-01-01

Full Text Available Horizontal velocity flow fields were measured by particle image velocimetry for a stirred vessel with baffles and two helical coils for enlargement of heat transfer area. The investigation was carried out in a cylindrical vessel with flat base and two different stirrers (radial-flow Rushton turbine and axial-flow propeller stirrer. Combined velocity plots for flow fields at different locations are presented. It was found that helical coils change the flow pattern significantly. Measurements for the axial-flow Rushton turbine showed a strong deflection by the coils, leading to a mainly tangential flow pattern. Behind baffles large areas of unused heat transfer area were found. First results for the axial-flow propeller reveal an extensive absence of fluid movement in the horizontal plane. Improved design considerations for enhanced heat transfer by more compatible equipment compilation are proposed.

Nuclear reactor pressure vessel-specific flaw distribution development

International Nuclear Information System (INIS)

Rosinski, S.T.

1992-01-01

Vessel integrity predictions performed through fracture mechanics analysis of a pressurized thermal shock event have been shown to be significantly sensitive to the overall flaw distribution input. It has also been shown that modem vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of US vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired US PWR vessel inspection data sets. Throughout the program, new insight was obtained into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. For example, the potential application of a vessel-specific flaw distribution now provides at least one method by which a vessel-specific reference flaw size applicable to pressure-temperature limit curves determination can be estimated. This paper will discuss the development and application of the methodology and the impact to future vessel integrity analyses

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-07-01

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

Dynamic response of the JT-60 vacuum vessel under the electromagnetic forces

International Nuclear Information System (INIS)

Takatsu, H.; Shimizu, M.; Ohta, M.

1982-01-01

Dynamic response analyses of the JAERI Tokamak 60 (JT-60) vacuum vessel were carried out under three kinds of saddle-like electromagnetic forces. In the analysis, the dynamic response of the bellows was obtained by dividing it into three components; the first, caused by the forced deflection due to the displacement of an adjacent rigid ring; the second, caused by inertia force; and the third, caused by a saddle-like electromagnetic force. Eigenvalue analyses showed that the 20th mode is a typical rotation mode of the rigid ring around the major radius with a natural frequency of 46.3 Hz. From the results of the dynamic response analyses, the maximum displacement response of the rigid ring was 3.1 mm and remarkable dynamic response was observed in the case of plasma disruption with a time constant of 1 ms. In cases of start-up of the plasma current and plasma disruption with a time constant of 50 ms, the rigid ring vibrates quasi-statically. It is clear that the dynamic behavior of the vacuum vessel is governed mainly by the saddle-like electromagnetic force, with a smaller effect of the inverse saddle-like electromagnetic force on the dynamic response of the vacuum vessel. (orig.)

Lymphatic Vessel Density as Prognostic Factor in Breast Carcinoma: Relation to Clinico pathologic Parameters

International Nuclear Information System (INIS)

El-Gendi, S.; Abdel-Hadi, M.

2009-01-01

Angiogenesis and lymphangiogenesis are essential for breast cancer growth and progression. This study aimed at investigating lymphatic micro vessel density (LVD) and microvessel density (MVD) as prognostic markers in breast carcinoma. Forty breast carcinomas were immuno stained for D2-40, CD31 and VEGF. Median lymphatic and blood micro vessel densities, as well as VEGF expression, were related to each other and to clinico pathologic parameters including lymph node (Ln) status. The efficacy of haematoxylin and eosin (H and E) in detecting lymphatic vessel invasion (LVI) compared to D2-40 immunostaining was also investigated. D2-40 stained normal lymphatic endothelium and myoepithelial cells, but with different staining patterns. D2-40 LVD related significantly to CD31 counts (r=0.470; p=0.002), and LN metastasis (Mann-Whitney U=101.500; p=0.043); however, it did not relate to age, tumor grade, tumor size or LVI. D2-40 identified LVI in 3 more cases (7.5%) than those detected by H and E. VEGF was expressed in 85% of cases, and was significantly related to CD31 and D2-40 counts (p=0.033 and 0.007, respectively). In conclusion, D2-40 LVD showed a significant association with LN metastasis, and can be considered to segregate patients with positive from those with negative LNs. D2-40 enhances the detection of LVI relative to H and E staining reflecting a potential for lymphatic metastatic spread and possible poor prognosis

Automated detection of kinks from blood vessels for optic cup segmentation in retinal images

Science.gov (United States)

Wong, D. W. K.; Liu, J.; Lim, J. H.; Li, H.; Wong, T. Y.

2009-02-01

The accurate localization of the optic cup in retinal images is important to assess the cup to disc ratio (CDR) for glaucoma screening and management. Glaucoma is physiologically assessed by the increased excavation of the optic cup within the optic nerve head, also known as the optic disc. The CDR is thus an important indicator of risk and severity of glaucoma. In this paper, we propose a method of determining the cup boundary using non-stereographic retinal images by the automatic detection of a morphological feature within the optic disc known as kinks. Kinks are defined as the bendings of small vessels as they traverse from the disc to the cup, providing physiological validation for the cup boundary. To detect kinks, localized patches are first generated from a preliminary cup boundary obtained via level set. Features obtained using edge detection and wavelet transform are combined using a statistical approach rule to identify likely vessel edges. The kinks are then obtained automatically by analyzing the detected vessel edges for angular changes, and these kinks are subsequently used to obtain the cup boundary. A set of retinal images from the Singapore Eye Research Institute was obtained to assess the performance of the method, with each image being clinically graded for the CDR. From experiments, when kinks were used, the error on the CDR was reduced to less than 0.1 CDR units relative to the clinical CDR, which is within the intra-observer variability of 0.2 CDR units.

[A study concerning how much weight schoolchildren carry in their bags, involving four schools in the metropolitan area of Buenos Aires, Argentina].

Science.gov (United States)

Laíño, Fernando A; Santa María, Claudio J; Bazán, Nelio E; Mainero, Daniel D

2013-01-01

Determining the weight children carry in their bags to school (absolute and relative values) and the distance walked during home-school routes, involving students from four schools in the metropolitan area of Buenos Aires. The study involved 751 primary (4th to 6th grades) and secondary (1st to 3rd years) level students who were attending three private schools and one public one. Body and bag weights were measured and the children were asked about the distance (in blocks) they walked from school to home. The study involved a descriptive analysis and contrasted the students by gender, educational level, type of school and grade or year. Possible associations between variables were ascertained. The group was divided into those carrying bags weighing less than 10% of their body weight and those who carrying 10% (considered a critical value) and more; frequencies were calculated by the type of bag being used. 68% of the sample were carrying 10% or more of their body weight (P42=10.13%): 66% in male (P44=10.12%) and 60% in female children (P40=10.2%). Private school students carried more weight than public school children (p<0.05) and younger students carried a greater weight than older students (p<0.05) in both educational levels. Most children were carrying relative weights well above that recommended and female students were most affected. Younger students carried higher absolute and relative weights.

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

High efficiency algorithm for 3D transient thermo-elasto-plastic contact problem in reactor pressure vessel sealing system

International Nuclear Information System (INIS)

Xu Mingyu; Lin Tengjiao; Li Runfang; Du Xuesong; Li Shuian; Yang Yu

2005-01-01

There are some complex operating cases such as high temperature and high pressure during the operating process of nuclear reactor pressure vessel. It is necessary to carry out mechanical analysis and experimental investigation for its sealing ability. On the basis of the self-developed program for 3-D transient sealing analysis for nuclear reactor pressure vessel, some specific measures are presented to enhance the calculation efficiency in several aspects such as the non-linear solution of elasto-plastic problem, the mixed solution algorithm for contact problem as well as contract heat transfer problem and linear equation set solver. The 3-D transient sealing analysis program is amended and complemented, with which the sealing analysis result of the pressure vessel model can be obtained. The calculation results have good regularity and the calculation efficiency is twice more than before. (authors)

Progress in the design and R and D of the ITER In-Vessel Viewing and Metrology System (IVVS)

Energy Technology Data Exchange (ETDEWEB)

Dubus, Gregory, E-mail: gregory.dubus@f4e.europa.eu [Fusion for Energy, c/ Josep Pla, n°2 – Torres Diagonal Litoral – Edificio B3, 08019 Barcelona (Spain); Puiu, Adrian; Bates, Philip; Damiani, Carlo [Fusion for Energy, c/ Josep Pla, n°2 – Torres Diagonal Litoral – Edificio B3, 08019 Barcelona (Spain); Reichle, Roger; Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2014-10-15

The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at performing inspections as well as providing information related to the erosion of in-vessel components, which in turn is related to the amount of mobilised dust present in the Vacuum Vessel. Periodically or on request, the IVVS scanning probes will be deployed into the Vacuum Vessel in order to acquire both visual and metrological data on plasma facing components (blanket, divertor, heating/diagnostic plugs, and test blanket modules). Recent design changes made to the six IVVS port extensions implied the need for a substantial redesign of the IVVS integrated concept – including the scanning probe and its deployment system – in order to bring it to the level of maturity suitable for the Conceptual Design Review. This paper gives an overview of the concept design for IVVS as well as of the various engineering analyses and R and D activities carried out in support to this design: neutronic, seismic and electromagnetic analyses, probe actuation validation under environmental conditions.

Investigation of the templets cut out of the Kozloduy Unit 2 reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Klausnitzer, E; Leitz, C; Rieg, C Y [Electricite de France (EDF), 75 - Paris (France); Kryukov, A [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

1994-12-31

Instrumented impact tests were carried out on sub-size specimens of base and weld metal. Using correlation dependencies, values of ductile-to-brittle transition temperature were determined before and after annealing for base and weld metal corresponding to standard specimen tests. The experimental results were compared to a prediction of the extent of radiation-induced embrittlement of Kozloduy Unit 2 pressure vessel materials. It was confirmed that radiation-induced embrittlement of the base metal does not impose any limitation on the radiation-limited lifetime of the pressure vessel. The predicted increase of the ductile-to-brittle transition temperature of weld metal as a result of irradiation (about 165 C) is practically equal to the experimental result (162 C). The ductile-to-brittle transition temperature of the weld metal was recovered by no less than 85 %.

Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel. Pt. 2

International Nuclear Information System (INIS)

Rose, P.W.

1987-12-01

In previous experiments, done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel, the instrument tubes support structure built into the vault was not included. It consists of a number of grids made up of fairly massive steel girders. These have now been added to the model and experiments performed using water to simulate molten core debris assumed to have fallen on to the vault floor and high-pressure air to simulate the discharge of steam or gas from the assumed breach at the bottom of the pressure vessel. The results show that the tubes support structure considerably reduces the carry-over of liquid via the vault access shafts. (author)

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

International Nuclear Information System (INIS)

Steinwarz, Wolfgang; Bounin, Dieter

2014-01-01

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

Graywater Discharges from Vessels

Science.gov (United States)

2011-11-01

metals (e.g., cadmium, chromium, lead, copper , zinc, silver, nickel, and mercury), solids, and nutrients (USEPA, 2008b; USEPA 2010). Wastewater from... flotation ), and disinfection (using ultraviolet light) as compared to traditional Type II MSDs that use either simple maceration and chlorination, or...Coliform Naval Vessels Oceanographic Vessels Small Cruise Ships 25a Vendor 2 Hamann AG Biological Treatment with Dissolved Air Flotation and

Reactor vessel decommissioning project. Final report

International Nuclear Information System (INIS)

Schoonen, D.H.

1984-09-01

This report describes a reactor vessel decommissioning project; it documents and explains the project objectives, scope, performance results, and sodium removal process. The project was successfully completed in FY-1983, within budget and without significant problems or adverse impact on the environment. Waste generated by the operation included the reactor vessel, drained sodium, and liquid, solid, and gaseous wastes which were significantly less than project estimates. Personnel radiation exposures were minimized, such that the project total was one-half the predicted exposure level. Except for the sodium removed, the material remaining in the reactor vessel is essentially the same as when the vessel arrived for processing

Electrical discharge machining for vessel sample removal

International Nuclear Information System (INIS)

Litka, T.J.

1993-01-01

Due to aging-related problems or essential metallurgy information (plant-life extension or decommissioning) of nuclear plants, sample removal from vessels may be required as part of an examination. Vessel or cladding samples with cracks may be removed to determine the cause of cracking. Vessel weld samples may be removed to determine the weld metallurgy. In all cases, an engineering analysis must be done prior to sample removal to determine the vessel's integrity upon sample removal. Electrical discharge machining (EDM) is being used for in-vessel nuclear power plant vessel sampling. Machining operations in reactor coolant system (RCS) components must be accomplished while collecting machining chips that could cause damage if they become part of the flow stream. The debris from EDM is a fine talclike particulate (no chips), which can be collected by flushing and filtration

State Gun Law Environment and Youth Gun Carrying in the United States.

Science.gov (United States)

Xuan, Ziming; Hemenway, David

2015-11-01

Gun violence and injuries pose a substantial threat to children and youth in the United States. Existing evidence points to the need for interventions and policies for keeping guns out of the hands of children and youth. (1) To examine the association between state gun law environment and youth gun carrying in the United States, and (2) to determine whether adult gun ownership mediates this association. This was a repeated cross-sectional observational study design with 3 years of data on youth gun carrying from US states. The Youth Risk Behavior Survey comprises data of representative samples of students in grades 9 to 12 from biennial years of 2007, 2009, and 2011. We hypothesized that states with more restrictive gun laws have lower rates of youth gun carrying, and this association is mediated by adult gun ownership. State gun law environment as measured by state gun law score. Youth gun carrying was defined as having carried a gun on at least 1 day during the 30 days before the survey. In the fully adjusted model, a 10-point increase in the state gun law score, which represented a more restrictive gun law environment, was associated with a 9% decrease in the odds of youth gun carrying (adjusted odds ratio [AOR], 0.91 [95% CI, 0.86-0.96]). Adult gun ownership mediated the association between state gun law score and youth gun carrying (AOR, 0.94 [ 95% CI, 0.86-1.01], with 29% attenuation of the regression coefficient from -0.09 to -0.07 based on bootstrap resampling). More restrictive overall gun control policies are associated with a reduced likelihood of youth gun carrying. These findings are relevant to gun policy debates about the critical importance of strengthening overall gun law environment to prevent youth gun carrying.

Nuclear power plant pressure vessels. Inservice inspections

International Nuclear Information System (INIS)

1995-01-01

The requirements for the planning and reporting of inservice inspections of nuclear power plant pressure vessels are presented. The guide specifically applies to inservice inspections of Safety class 1 and 2 nuclear power plant pressure vessels, piping, pumps and valves plus their supports and reactor pressure vessel internals by non- destructive examination methods (NDE). Inservice inspections according to the Pressure Vessel Degree (549/73) are discussed separately in the guide YVL 3.0. (4 refs.)

46 CFR 15.405 - Familiarity with vessel characteristics.

Science.gov (United States)

2010-10-01

... 46 Shipping 1 2010-10-01 2010-10-01 false Familiarity with vessel characteristics. 15.405 Section... MANNING REQUIREMENTS Manning Requirements; All Vessels § 15.405 Familiarity with vessel characteristics. Each credentialed individual must become familiar with the relevant characteristics of the vessel on...

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Carry

DEFF Research Database (Denmark)

Koijen, Ralph S.J.; Moskowitz, Tobias; Pedersen, Lasse Heje

2018-01-01

-sectionally and in time series for a host of different asset classes, including global equities, global bonds, commodities, US Treasuries, credit, and options. Carry is not explained by known predictors of returns from these asset classes, and it captures many of these predictors, providing a unifying framework...... for return predictability. We reject a generalized version of Uncovered Interest Parity and the Expectations Hypothesis in favor of models with varying risk premia, in which carry strategies are commonly exposed to global recession, liquidity, and volatility risks, though none fully explains carry’s premium....

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Design of the ITER vacuum vessel

International Nuclear Information System (INIS)

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

1995-01-01

The ITER vacuum vessel is a major safety barrier and must support electromagnetic loads during plasma disruptions and vertical displacement events (VDE) and withstand plausible accidents without losing confinement.The vacuum vessel has a double wall structure to provide structural and electrical continuity in the toroidal direction. The inner and outer shells and poloidal stiffening ribs between them are joined by welding, which gives the vessel the required mechanical strength. The space between the shells will be filled with steel balls and plate inserts to provide additional nuclear shielding. Water flowing in this space is required to remove nuclear heat deposition, which is 0.2-2.5% of the total fusion power. The minor and major radii of the tokamak are 3.9 m and 13 m respectively, and the overall height is 15 m. The total thickness of the vessel wall structure is 0.4-0.7 m.The inboard and outboard blanket segments are supported from the vacuum vessel. The support structure is required to withstand a large total vertical force of 200-300 MN due to VDE and to allow for differential thermal expansion.The first candidate for the vacuum vessel material is Inconel 625, due to its higher electric resistivity and higher yield strength, even at high temperatures. Type 316 stainless steel is also considered a vacuum vessel material candidate, owing to its large database and because it is supported by more conventional fabrication technology. (orig.)

No evidence for an open vessel effect in centrifuge-based vulnerability curves of a long-vesselled liana (Vitis vinifera).

Science.gov (United States)

Jacobsen, Anna L; Pratt, R Brandon

2012-06-01

Vulnerability to cavitation curves are used to estimate xylem cavitation resistance and can be constructed using multiple techniques. It was recently suggested that a technique that relies on centrifugal force to generate negative xylem pressures may be susceptible to an open vessel artifact in long-vesselled species. Here, we used custom centrifuge rotors to measure different sample lengths of 1-yr-old stems of grapevine to examine the influence of open vessels on vulnerability curves, thus testing the hypothesized open vessel artifact. These curves were compared with a dehydration-based vulnerability curve. Although samples differed significantly in the number of open vessels, there was no difference in the vulnerability to cavitation measured on 0.14- and 0.271-m-long samples of Vitis vinifera. Dehydration and centrifuge-based curves showed a similar pattern of declining xylem-specific hydraulic conductivity (K(s)) with declining water potential. The percentage loss in hydraulic conductivity (PLC) differed between dehydration and centrifuge curves and it was determined that grapevine is susceptible to errors in estimating maximum K(s) during dehydration because of the development of vessel blockages. Our results from a long-vesselled liana do not support the open vessel artifact hypothesis. © 2012 The Authors. New Phytologist © 2012 New Phytologist Trust.

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

International Nuclear Information System (INIS)

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

1975-11-01

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

Numerical analysis of coolant mixing in the pressure vessel of WWER-440 type nuclear reactors

International Nuclear Information System (INIS)

Boros, I.; Aszodi, A.

2003-01-01

The precise description of the coolant mixing processes taking place in the reactor pressure vessel (RPV) of pressurized water nuclear reactors has an essential importance during power operation, as well as in case of incidental or accidental conditions. In this paper the detailed CFD model of the pressure vessel of a WWER-440 type reactor and calculations performed with this RPV model are presented. The CFD model of the pressure vessel contains all the important internal structural elements of the RPV. Sensitivity study on the effect of these elements was also carried out. Both steady-state and transient calculation were performed using the CFD code CFX-5.5.1. The results of the steady-state calculations give the so called mixing factors, i.e. the effect of each single primary loop at the core inlet. The mixing factors can be given for nominal circumstances (i.e. all main coolant pumps are working) or in case of less than six working MCPs. In order to validate the model the calculated mixing factors are compared with the values measured in the Paks NPP (Authors)

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)

TPX vacuum vessel transient thermal and stress conditions

International Nuclear Information System (INIS)

Feldshteyn, Y.; Dinkevich, S.; Feng, T.; Majumder, D.

1995-01-01

The TPX vacuum vessel provides the vacuum boundary for the plasma and the mechanical support for the internal components. Another function of the vacuum vessel is to contain neutron shielding water in the double wall space during normal operation. This double wall space serves as a heat reservoir for the entire vacuum vessel during bakeout. The vacuum vessel and the internal components are subjected to thermal stresses induced by a nonuniform temperature distribution within the structure during bakeout. A successful Conceptual Design Review in March 1993 has established superheated steam as the heating source of the vacuum vessel. A transient bakeout mode of the vacuum vessel and in-vessel components has been analyzed to evaluate transient period duration, proper temperature level, actual thermal stresses and performance of the steam equipment. Thermally, the vacuum vessel structure may be considered as an adiabatic system because it is perfectly insulated by the strong surrounding vacuum and multiple layers of superinsulation. Important aspects of the analysis are described herein

In service inspection of SUPERPHENIX 1 vessels: MIR

International Nuclear Information System (INIS)

Asty, M.; Viard, J.; Lerat, B.; Saglio, R.

1985-01-01

Although no in-service inspection constraints were imposed on the Phenix vessels, the Safety Authorities asked that the design of SUPERPHENIX 1 makes it possible to monitor throughout the lifetime of the reactor, surface and internal defects on the main vessel. A pool design and the presence of heat baffles inside the main vessel make access from the inside of the vessel impossible. Thus, an inspection can only be performed from the outside of the main vessel: the distance between the walls of the main and safety vessels is such that an inspection device can be introduced into the corresponding space. As the design of the reactor precludes radiographic inspection, the method which was selected for monitoring internal defects in the main vessel is ultrasonics. However, the anisotropic structure of austenitic stainless steel welds limits the performance of this technique. The authors present the in-service inspection device, MIR, which has been specially developed for the visual and ultrasonic examination of SUPERPHENIX 1 vessels

Final report for the 2nd Ex-Vessel Neutron Dosimetry Installations and Evaluations for Yonggwang Unit 2 Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

2008-01-15

This report describes a neutron fluence assessment performed for the Yonggwang Unit 2 pressure vessel beltline region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. During Cycle 16 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 2 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

Final report for the 1st ex-vessel neutron dosimetry installations and evaluations for Kori unit 2 reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

2006-11-15

This report describes a neutron fluence assessment performed for the Kori unit 2 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 20 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 20.

Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Li, Nam Jin; Hong, Joon Wha

2007-01-15

This report describes a neutron fluence assessment performed for the Yonggwang Unit 1 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 16 of reactor operation, 2nd Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 1 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

2008-01-15

This report describes a neutron fluence assessment performed for the Kori Unit 2 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 21 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 21.

Final report for the 1st ex-vessel neutron dosimetry installation and evaluations for Kori unit 4 reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

2006-11-15

This report describes a neutron fluence assessment performed for the Kori unit 4 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 16 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 4 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 16.