Numerical study on interaction of local air cooler with stratified hydrogen cloud in a large vessel

Energy Technology Data Exchange (ETDEWEB)

Liang, Z. [Atomic Energy of Canada Limited, Chalk River Laboratories, ON K0J 1J0 (Canada); Andreani, M. [Laboratory for Thermal-Hydraulics, Paul Scherrer Institut, 5232 Villigen (Switzerland)

2012-07-01

Within the framework of the ERCOSAM project, planning calculations are performed to examine sensitivity parameters that can affect the break-up (erosion) of a helium layer by mitigation devices (i.e., cooler, spray, or Passive Autocatalytic Recombiner - PAR). This paper reports the GOTHIC analysis results for the cooler tests to be performed in the PANDA facility. The cooler elevation and geometry, helium layer thickness, steam distribution in the vessel, and the vessel geometry (inter-connected multi-compartments versus a single volume) on the erosion process as well as the cooling capacity are studied. This analysis is valuable because only a limited number of conditions will be examined in the planned experiments. The study provides a useful understanding of the interaction of a cooler with a stratified atmosphere. (authors)

Hydrogen Gas Retention and Release from WTP Vessels: Summary of Preliminary Studies

Energy Technology Data Exchange (ETDEWEB)

Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bontha, Jagannadha R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Daniel, Richard C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mahoney, Lenna A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rassat, Scot D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wells, Beric E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bao, Jie [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Boeringa, Gregory K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buchmiller, William C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burns, Carolyn A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chun, Jaehun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Karri, Naveen K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Li, Huidong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tran, Diana N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-07-01

The Hanford Waste Treatment and Immobilization Plant (WTP) is currently being designed and constructed to pretreat and vitrify a large portion of the waste in the 177 underground waste storage tanks at the Hanford Site. A number of technical issues related to the design of the pretreatment facility (PTF) of the WTP have been identified. These issues must be resolved prior to the U.S. Department of Energy (DOE) Office of River Protection (ORP) reaching a decision to proceed with engineering, procurement, and construction activities for the PTF. One of the issues is Technical Issue T1 - Hydrogen Gas Release from Vessels (hereafter referred to as T1). The focus of T1 is identifying controls for hydrogen release and completing any testing required to close the technical issue. In advance of selecting specific controls for hydrogen gas safety, a number of preliminary technical studies were initiated to support anticipated future testing and to improve the understanding of hydrogen gas generation, retention, and release within PTF vessels. These activities supported the development of a plan defining an overall strategy and approach for addressing T1 and achieving technical endpoints identified for T1. Preliminary studies also supported the development of a test plan for conducting testing and analysis to support closing T1. Both of these plans were developed in advance of selecting specific controls, and in the course of working on T1 it was decided that the testing and analysis identified in the test plan were not immediately needed. However, planning activities and preliminary studies led to significant technical progress in a number of areas. This report summarizes the progress to date from the preliminary technical studies. The technical results in this report should not be used for WTP design or safety and hazards analyses and technical results are marked with the following statement: “Preliminary Technical Results for Planning – Not to be used for WTP Design

Models and Algorithms for Container Vessel Stowage Optimization

DEFF Research Database (Denmark)

Delgado-Ortegon, Alberto

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

Research program plan: reactor vessels. Volume 1

International Nuclear Information System (INIS)

Vagins, M.; Taboada, A.

1985-07-01

The ability of the licensing staff of the NRC to make decisions concerning the present and continuing safety of nuclear reactor pressure vessels under both normal and abnormal operating conditions is dependent upon the existence of verified analysis methods and a solid background of applicable experimental data. It is the role of this program to provide both the analytical methods and the experimental data needed. Specifically, this program develops fracture mechanics analysis methods and design criteria for predicting the stress levels and flaw sizes required for crack initiation, propagation, and arrest in LWR pressure vessels under all known and postulated operations conditions. To do this, not only must the methods be developed but they must be experimentally validated. Further, the materials data necessary for input to these analytical methods must be developed. Thus, in addition to methods development and large scale experimental verification this program also develops data to show that slow-load fracture toughness, rapid-load fracture toughness, and crack arrest toughness obtained from small laboratory specimens are truly representative of the toughness characteristics of the material behavior in pressure vessels in both the unirradiated and the irradiated conditions

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.

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.)

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.

Vessel Monitoring Systems Study. Volume I - Technical Analysis.

Science.gov (United States)

1980-09-01

In the Port and Tanker Safety Act of 1978 the U.S. Conress directed the Department of Transportation to performa a study on the desirability and feasibility of a shore-station system for monitoring vessels (including fishing vessels)offshore within t...

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.

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)

Reactor vessel pressure transient protection for pressurized water reactors

International Nuclear Information System (INIS)

Zech, G.

1978-09-01

During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

Computerized reactor pressure vessel materials information system

International Nuclear Information System (INIS)

Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

1980-10-01

A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

Bottom head failure program plan

International Nuclear Information System (INIS)

Meyer, R.O.

1989-01-01

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

MORPHOMETRIC STUDY OF THE AREOLAR SPACE BETWEEN THE GREAT VESSELS AND THE LUMBAR SPINE

Directory of Open Access Journals (Sweden)

Luis Marchi

2015-12-01

Full Text Available Objective : This work aims to study the areolar space anterior to the lumbar spine, and also the positioning of the large vessels focusing a lateral approach. Methods :This is a morphometric study of 108 cases based on T2 weighted-MRI images in the supine position. The following measurements were performed: lumbar and segmental lordosis; anteroposterior disc diameter; space between the disc/vertebral body and the vessels; bifurcation between the abdominal aorta and the common iliac veins confluence in relation to the lumbar level. Results :The areolar space with respect to the iliac veins, and with the vena cava increased cranially (p<0.001, starting from average 0.6mm at L4-L5 and reaching 8.4mm at L2, while the abdominal aorta showed no increase or decrease pattern across the different levels (p=0.135 ranging from 1.8 to 4.6mm. The diameter of the discs increased distally (p<0.01 as well as the lordosis (p<0.001. The disc diameter was 11% larger when compared to the adjacent vertebral bodies (p<0.001 and that resulted in a smaller distance of the vessels in the disc level than in the level of the adjacent vertebral bodies (p<0.001. The aortic bifurcation was generally ahead of L4 (52% and less frequently at L3-L4 (28% and L4-L5 (18%. The confluence of the veins was usually at the L4-L5 level (38% and at L5 (37%, and less frequently at L4 (26%. Conclusions : There is an identifiable plane between the great vessels and the lumbar spine which is particularly narrow in its distal portion. It is theoretically feasible to reach this plan, handle the anterior complex disc/ALL and protect the great vessels by lateral approach, however, it is challenging.

Case study for one-piece removal method of reactor vessel of nuclear ship 'Mutsu'

International Nuclear Information System (INIS)

Nagane, Satoru; Kitahara, Katsumi; Yoshikawa, Seiji; Miyasaka, Yasuhiko; Fukumura, Nobuo; Nisizawa, Ichiou

2010-01-01

A reactor installed at the center part of the nuclear ship 'Mutsu' has been stored safely and exhibited in a reactor room building since 1996. The reactor vessel and its internals are key components because of main radioactive wastes for the reasonable decommissioning plan in the future. This report describes the one-piece removal method as the one package of the reactor vessel with its internals intact with a shipping container or additional shields. The reactor vessel package (Max.100ton) will be classified acceptable for burial at the low level radioactive waste (LLW), which will be buried at a LLW pit facility under waste disposal regulations. And also, the package will be classified as an IP-2-equivalent package according to the requirement for Shipments and Packagings. (author)

Experimental studies of oxidic molten corium-vessel steel interaction

International Nuclear Information System (INIS)

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

2001-01-01

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

Experimental studies of oxidic molten corium-vessel steel interaction

Energy Technology Data Exchange (ETDEWEB)

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

2001-12-01

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

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.

Arctic research vessel design would expand science prospects

Science.gov (United States)

Elsner, Robert; Kristensen, Dirk

The U.S. polar marine science community has long declared the need for an arctic research vessel dedicated to advancing the study of northern ice-dominated seas. Planning for such a vessel began 2 decades ago, but competition for funding has prevented construction. A new design program is underway, and it shows promise of opening up exciting possibilities for new research initiatives in arctic marine science.With its latest design, the Arctic Research Vessel (ARV) has grown to a size and capability that will make it the first U.S. academic research vessel able to provide access to the Arctic Ocean. This ship would open a vast arena for new studies in the least known of the world's seas. These studies promise to rank high in national priority because of the importance of the Arctic Ocean as a source of data relating to global climate change. Other issues that demand attention in the Arctic include its contributions to the world's heat budget, the climate history buried in its sediments, pollution monitoring, and the influence of arctic conditions on marine renewable resources.

Careful treatment planning enables safe ablation of liver tumors adjacent to major blood vessels by percutaneous irreversible electroporation (IRE).

Science.gov (United States)

Kos, Bor; Voigt, Peter; Miklavcic, Damijan; Moche, Michael

2015-09-01

Irreversible electroporation (IRE) is a tissue ablation method, which relies on the phenomenon of electroporation. When cells are exposed to a sufficiently electric field, the plasma membrane is disrupted and cells undergo an apoptotic or necrotic cell death. Although heating effects are known IRE is considered as non-thermal ablation technique and is currently applied to treat tumors in locations where thermal ablation techniques are contraindicated. The manufacturer of the only commercially available pulse generator for IRE recommends a voltage-to-distance ratio of 1500 to 1700 V/cm for treating tumors in the liver. However, major blood vessels can influence the electric field distribution. We present a method for treatment planning of IRE which takes the influence of blood vessels on the electric field into account; this is illustrated on a treatment of 48-year-old patient with a metastasis near the remaining hepatic vein after a right side hemi-hepatectomy. Output of the numerical treatment planning method shows that a 19.9 cm3 irreversible electroporation lesion was generated and the whole tumor was covered with at least 900 V/cm. This compares well with the volume of the hypodense lesion seen in contrast enhanced CT images taken after the IRE treatment. A significant temperature raise occurs near the electrodes. However, the hepatic vein remains open after the treatment without evidence of tumor recurrence after 6 months. Treatment planning using accurate computer models was recognized as important for electrochemotherapy and irreversible electroporation. An important finding of this study was, that the surface of the electrodes heat up significantly. Therefore the clinical user should generally avoid placing the electrodes less than 4 mm away from risk structures when following recommendations of the manufacturer.

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)

Report on Workshop "Planning of Future Science in the Polar Ocean Study with Cooperation among Study Groups"

Directory of Open Access Journals (Sweden)

Mitsuo Fukuchi

2001-03-01

Full Text Available A workshop on "Planning of Future Science in the Polar Ocean Study with Cooperation among Study Groups" was held on November 1,2000,at the National Institute of Polar Research with 21 participants. In this workshop, a plan to charter a research vessel other than "Shirase" was introduced and a science plan using the chartered research vessel by 43rd Japanese Antarctic Research Expedition was discussed. This study is going to be conducted in the sea ice area around 140-150°E in mid-summer (February 2002, when biological production becomes active in the Antarctic Ocean. Oceanographic observations using "Shirase" are difficult to conduct in this season since she supports a wide range of summer operations around Syowa Station. The relationships between biological production and greenhouse effect gas production and the vertical transport of organic materials from the surface to deep ocean will be the focus of this study. At this stage, one deputy leader and three members of JARE, and 25-26 other scientists including graduate students and foreign scientists, will participate in the field observations using the chartered vessel. The members of JARE will conduct a project science program of the VI Phase of JARE, while the other participants will do part of the science program "Antarctic Ocean in Earth System". Since further observations for several years after the summer of 2002 will be required to understand the role of the Antarctic Ocean in global climate change, we have applied for a Grant-in-Aid for Scientific Research for the next project, which will start from 2001,to the Ministry of Education, Science, Sports and Culture of Japan. The proposal was discussed in detail in this workshop.

Visits to Australia by nuclear powered or armed vessels: contingency planning for the accidental release of ionizing radiation

International Nuclear Information System (INIS)

1989-01-01

The report refers to the adequacy of current contingency planning by the Australian Federal and Senate authorities to deal with the accidental release of ionizating radiation from visiting nuclear powered or armed vessels in Australian waters and ports. Much of the material was obtained in response to questions put in writing by the Senate Standing Committee to the Department of Defence, ANSTO and others. In addition, the report contains relevant information from Commonwealth documents as well as the Committee findings and recommendations. Issues considered include: types of visiting nuclear powered vessels, accident likelihood and consequences, differences between naval and land-based reactors, safety records. The persons or organizations who made submissions or appeared in all public hearings are listed in the appendixes, along with all visits to Australian ports by nuclear powered warships from 1976 to 1988

Laparoscopic prototype for optical sealing of renal blood vessels

Science.gov (United States)

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

2017-02-01

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

Conceptual design for Japan Sodium-Cooled Fast Reactor. (4) Developmental study of steel plate reinforced concrete containment vessel for JSFR

International Nuclear Information System (INIS)

Hosoya, Takusaburo; Negishi, Kazuo; Satoh, Kenichiro; Somaki, Takahiro; Matsuo, Ippei; Shimizu, Katsusuke

2009-01-01

An innovative containment vessel, namely Steel plate reinforced Concrete Containment Vessel (SCCV) is developed for Japan Sodium-Cooled Fast Reactor (JSFR). Reducing plant construction cost is one of the most important issues for commercialization of fast reactors. This study investigated construction issues including the building structure and the construction method as well as design issues in terms of the applicability of SCCV to fast reactors. An experimental study including loading and/or heating tests has been carried out to investigate the fundamental structural features, which would be provided to develop methodology to evaluate the feasibility of SCCV under the severe conditions. In this paper, the test plan is described as well as the first test results. (author)

Optimization of reactor pressure vessel internals segmentation in Korea

Energy Technology Data Exchange (ETDEWEB)

Lee, Byung-Sik [Dankook Univ., Chungnam (Korea, Republic of). Dept. of Nuclear Engineering

2017-11-15

One of the most challenging tasks during plant decommissioning is the removal of highly radioactive internal components from the reactor pressure vessel (RPV). For RPV internals dismantling, it is essential that all activities are thoroughly planned and discussed in the early stage of the decommissioning project. One of the key activities in the detailed planning is to prepare the segmentation and packaging plan that describes the sequential steps required to segment, separate, and package each individual component of RPV, based on an activation analysis and component characterization study.

An experimental system for the study of ultrasound exposure of isolated blood vessels

International Nuclear Information System (INIS)

Tokarczyk, Anna; Rivens, Ian; Symonds-Tayler, Richard; Ter Haar, Gail; Van Bavel, E

2013-01-01

An experimental system designed for the study of the effects of diagnostic or therapeutic ultrasound exposure on isolated blood vessels in the presence or absence of intraluminal contrast agent is described. The system comprised several components. A microscope was used to monitor vessel size (and thus vessel functionality), and potential leakage of intraluminal 70 kDa FITC-dextran fluorescence marker. A vessel chamber allowed the mounting of an isolated vessel whilst maintaining its viability, with pressure regulation for the control of intraluminal pressure and induction of flow for the infusion of contrast microbubbles. A fibre-optic hydrophone sensor mounted on the vessel chamber using a micromanipulator allowed pre-exposure targeting of the vessel to within 150 µm, and monitoring of acoustic cavitation emissions during exposures. Acoustic cavitation was also detected using changes in the ultrasound drive voltage and by detection of audible emissions using a submerged microphone. The suitability of this system for studying effects in the isolated vessel model has been demonstrated using a pilot study of 6 sham exposed and 18 high intensity focused ultrasound exposed vessels, with or without intraluminal contrast agent (SonoVue) within the vessels. (paper)

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

International Nuclear Information System (INIS)

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

1992-01-01

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

A dynamic simulation to study NET in-vessel handling operations

International Nuclear Information System (INIS)

Fung, P.T.F.

1989-01-01

The inspection, maintenance and repair of the Next European Torus (NET) fusion machine will require the extensive use of remote handling equipment to minimise the human exposure to the high radiation environment. The use of efficient manipulators will reduce the NET downtime by reducing the preparation time for entry into the controlled area and by performing the task with reasonable area and by performing the task with reasonable dexterity and speed, consistent with safety. A high fidelity simulation is a valuable tool to assist in the manipulator design, operations, trajectory planning, parameter optimisation and system verification. A manipulator simulation package called ASAD was originally developed by Spar for space manipulator applications. It is now being adapted to simulate the in-Vessel HandlingUnit for the NET program. This terestrial version of ASAD has been name ASAD - T. Spar, through the services of the Canadian Fusion Fuels Technology Project, is under contract to the NET program for the performance of this activity. This paper describes the capabillities and underlying assumptions of ASAD - T, aling with description of the simulation development of the NET in-vessel manipulator. (author). 4 refs.; 7 figs

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.

Developments of high-performance moderator vessel for JRR-3 cold neutron source

International Nuclear Information System (INIS)

Arai, Masaji; Tamura, Itaru; Hazawa, Tomoya

2015-05-01

The cold neutron source (CNS) facility converts thermal neutrons into cold neutrons to moderate neutrons with liquid hydrogen. The cold neutron beam at Japan Research Reactor No. 3 (JRR-3) is led to the beam experimental devices in the beam hall through neutron guide tubes. High intensities of the cold neutron beam are always demanded for increasing the experimental effectiveness and accuracy. In the Department of Research Reactor and Tandem Accelerator, developments of high-performance CNS moderator vessel that can produce cold neutron intensity about two times higher compared to the existing vessel have been performed in the second medium term plans. We compiled this report about the technological development to solve several problems with the design and manufacture of new vessel. In the present study, design strength evaluation, mockup test, simulation for thermo-fluid dynamics of the liquid hydrogen and strength evaluation of the different-material-bonding were studied. By these evaluation results, we verified that the developed new vessel can be applied to CNS moderator vessel of JRR-3. (author)

Baking results of KSTAR vacuum vessel

International Nuclear Information System (INIS)

Kim, S. T.; Kim, Y. J.; Kim, K. M.; Im, D. S.; Joung, N. Y.; Yang, H. L.; Kim, Y. S.; Kwon, M.

2009-01-01

The Korea Superconducting Tokamak Advanced Research (KSTAR) is an advanced superconducting tokamak designed to establish a scientific and technological basis for an attractive fusion reactor. The fusion energy in the tokamak device is released through fusion reactions of light atoms such as deuterium or helium in hot plasma state, of which temperature reaches several hundreds of millions Celsius. The high temperature plasma is created in the vacuum vessel that provides ultra high vacuum status. Accordingly, it is most important for the vacuum condition to keep clean not only inner space but also surface of the vacuum vessel to make high quality plasma. There are two methods planned to clean the wall surface of the KSTAR vacuum vessel. One is surface baking and the other is glow discharge cleaning (GDC). To bake the vacuum vessel, De-Ionized (DI) water is heated to 130 .deg. C and circulated in the passage between double walls of the vacuum vessel (VV) in order to bake the surface. The GDC operation uses hydrogen and inert gas discharges. In this paper, general configuration and brief introduction of the baking result will be reported

Baking results of KSTAR vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, S. T.; Kim, Y. J.; Kim, K. M.; Im, D. S.; Joung, N. Y.; Yang, H. L.; Kim, Y. S.; Kwon, M. [National Fusion Research Institute, Daejeon (Korea, Republic of)

2009-05-15

The Korea Superconducting Tokamak Advanced Research (KSTAR) is an advanced superconducting tokamak designed to establish a scientific and technological basis for an attractive fusion reactor. The fusion energy in the tokamak device is released through fusion reactions of light atoms such as deuterium or helium in hot plasma state, of which temperature reaches several hundreds of millions Celsius. The high temperature plasma is created in the vacuum vessel that provides ultra high vacuum status. Accordingly, it is most important for the vacuum condition to keep clean not only inner space but also surface of the vacuum vessel to make high quality plasma. There are two methods planned to clean the wall surface of the KSTAR vacuum vessel. One is surface baking and the other is glow discharge cleaning (GDC). To bake the vacuum vessel, De-Ionized (DI) water is heated to 130 .deg. C and circulated in the passage between double walls of the vacuum vessel (VV) in order to bake the surface. The GDC operation uses hydrogen and inert gas discharges. In this paper, general configuration and brief introduction of the baking result will be reported.

A study of reactor vessel integrity assessment

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Hoon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Kim, Jong Kyung; Shin, Chang Ho; Seo, Bo Kyun [Hanyang Univ., Seoul (Korea, Republic of)

1999-02-15

The fast neutron fluence at the Reactor Pressure Vessel(RPV) of KNGR designed for 60 years lifetime was calculated by full-scope Monte Carlo simulation for reactor vessel integrity assessment. KNGR core geometry was modeled on a three-dimensional representation of the one-sixteenth of the reactor in-vessel component. Each fuel assemblies were modeled explicitly, and each fuel pins were axially divided into 5 segments. The maximum flux of 4.3 x 10{sup 10} neutrons/cm{sup 2}. sec at the RPV was obtained by tallying neutrons crossing the beltline of inner surface of the RPV.

The case for using vessel-based systems to apply oil-spill dispersants

International Nuclear Information System (INIS)

Ross, S.

1998-01-01

Most emergency plans for dealing with marine oil spills include the use of chemical dispersants. This paper presents a comparison between the capabilities of aircraft-based dispersant application systems and vessel-based systems. The comparison was presented in terms of the logistics of treating offshore spills. Vessel-based systems have certain advantages in terms of their availability and cost. They have better spray control and accuracy and can dose thick slicks in one pass. However, this advantage is lost if the dispersant payload on the vessel is relatively small and the spill is located very far from the base of operations. Under certain conditions, vessel-based dispersant application systems can treat spills as quickly as aircraft-based systems. Most marine spills tend to occur in restricted waters near ports where dispersant stockpiles could be stored for ready use by vessels in the area. Development of a modern vessel-based, fire-monitor system for applying chemical dispersant onto marine oil spills was one of the recommendations emerging from the study. 32 refs., 6 tabs., 1 fig

The Planning, Licensing, Modifications, and Use of a Russian Vessel for Shipping Spent Nuclear Fuel by Sea in Support of the DOE RRRFR Program

International Nuclear Information System (INIS)

Tyacke, Michael; Bolshinsky, Igor; Tomczak, Wlodzimierz; Naletov, Sergey; Pichugin, Oleg

2001-01-01

The Russian Research Reactor Fuel Return (RRRFR) Program, under the U.S. Department of Energy's Global Threat Reduction Initiative, began returning Russian-supplied high-enriched uranium (HEU) spent nuclear fuel (SNF), stored at Russian-designed research reactors throughout the world, to Russia in January 2006. During the first years of making HEU SNF shipments, it became clear that the modes of transportation needed to be expanded from highway and railroad to include sea and air to meet the extremely aggressive commitment of completing the first series of shipments by the end of 2010. The first shipment using sea transport was made in October 2008 and used a non-Russian flagged vessel. The Russian government reluctantly allowed a one-time use of the foreign-owned vessel into their highly secured seaport, with the understanding that any future shipments would be made using a vessel owned and operated by a Russian company. ASPOL-Baltic of St. Petersburg, Russia, owns and operates a small fleet of vessels and has a history of shipping nuclear materials. ASPOL-Baltic's vessels were licensed for shipping nuclear materials; however, they were not licensed to transport SNF materials. After a thorough review of ASPOL Baltic's capabilities and detailed negotiations, it was agreed that a contract would be let with ASPOL-Baltic to license and refit their MCL Trader vessel for hauling SNF in support of the RRRFR Program. This effort was funded through a contract between the RRRFR Program, Idaho National Laboratory, and Radioactive Waste Management Plant of Swierk, Poland. This paper discusses planning, Russian and international maritime regulations and requirements, Russian authorities reviews and approvals, licensing, design, and modifications made to the vessel in preparation for SNF shipments. A brief summary of actual shipments using this vessel, experiences, and lessons learned also are described.

The Planning, Licensing, Modifications, and Use of a Russian Vessel for Shipping Spent Nuclear Fuel by Sea in Support of the DOE RRRFR Program

Energy Technology Data Exchange (ETDEWEB)

Michael Tyacke; Dr. Igor Bolshinsky; Wlodzimierz Tomczak; Sergey Naletov; Oleg Pichugin

2001-10-01

The Russian Research Reactor Fuel Return (RRRFR) Program, under the U.S. Department of Energy’s Global Threat Reduction Initiative, began returning Russian-supplied high-enriched uranium (HEU) spent nuclear fuel (SNF), stored at Russian-designed research reactors throughout the world, to Russia in January 2006. During the first years of making HEU SNF shipments, it became clear that the modes of transportation needed to be expanded from highway and railroad to include sea and air to meet the extremely aggressive commitment of completing the first series of shipments by the end of 2010. The first shipment using sea transport was made in October 2008 and used a non-Russian flagged vessel. The Russian government reluctantly allowed a one-time use of the foreign-owned vessel into their highly secured seaport, with the understanding that any future shipments would be made using a vessel owned and operated by a Russian company. ASPOL-Baltic of St. Petersburg, Russia, owns and operates a small fleet of vessels and has a history of shipping nuclear materials. ASPOL-Baltic’s vessels were licensed for shipping nuclear materials; however, they were not licensed to transport SNF materials. After a thorough review of ASPOL Baltic’s capabilities and detailed negotiations, it was agreed that a contract would be let with ASPOL-Baltic to license and refit their MCL Trader vessel for hauling SNF in support of the RRRFR Program. This effort was funded through a contract between the RRRFR Program, Idaho National Laboratory, and Radioactive Waste Management Plant of Swierk, Poland. This paper discusses planning, Russian and international maritime regulations and requirements, Russian authorities’ reviews and approvals, licensing, design, and modifications made to the vessel in preparation for SNF shipments. A brief summary of actual shipments using this vessel, experiences, and lessons learned also are described.

Optimization study on structural analyses for the J-PARC mercury target vessel

Science.gov (United States)

Guan, Wenhai; Wakai, Eiichi; Naoe, Takashi; Kogawa, Hiroyuki; Wakui, Takashi; Haga, Katsuhiro; Takada, Hiroshi; Futakawa, Masatoshi

2018-06-01

The spallation neutron source at the Japan Proton Accelerator Research Complex (J-PARC) mercury target vessel is used for various materials science studies, work is underway to achieve stable operation at 1 MW. This is very important for enhancing the structural integrity and durability of the target vessel, which is being developed for 1 MW operation. In the present study, to reduce thermal stress and relax stress concentrations more effectively in the existing target vessel in J-PARC, an optimization approach called the Taguchi method (TM) is applied to thermo-mechanical analysis. The ribs and their relative parameters, as well as the thickness of the mercury vessel and shrouds, were selected as important design parameters for this investigation. According to the analytical results of 18 model types designed using the TM, the optimal design was determined. It is characterized by discrete ribs and a thicker vessel wall than the current design. The maximum thermal stresses in the mercury vessel and the outer shroud were reduced by 14% and 15%, respectively. Furthermore, it was indicated that variations in rib width, left/right rib intervals, and shroud thickness could influence the maximum thermal stress performance. It is therefore concluded that the TM was useful for optimizing the structure of the target vessel and to reduce the thermal stress in a small number of calculation cases.

Probabilistic atlas based labeling of the cerebral vessel tree

Science.gov (United States)

Van de Giessen, Martijn; Janssen, Jasper P.; Brouwer, Patrick A.; Reiber, Johan H. C.; Lelieveldt, Boudewijn P. F.; Dijkstra, Jouke

2015-03-01

Preoperative imaging of the cerebral vessel tree is essential for planning therapy on intracranial stenoses and aneurysms. Usually, a magnetic resonance angiography (MRA) or computed tomography angiography (CTA) is acquired from which the cerebral vessel tree is segmented. Accurate analysis is helped by the labeling of the cerebral vessels, but labeling is non-trivial due to anatomical topological variability and missing branches due to acquisition issues. In recent literature, labeling the cerebral vasculature around the Circle of Willis has mainly been approached as a graph-based problem. The most successful method, however, requires the definition of all possible permutations of missing vessels, which limits application to subsets of the tree and ignores spatial information about the vessel locations. This research aims to perform labeling using probabilistic atlases that model spatial vessel and label likelihoods. A cerebral vessel tree is aligned to a probabilistic atlas and subsequently each vessel is labeled by computing the maximum label likelihood per segment from label-specific atlases. The proposed method was validated on 25 segmented cerebral vessel trees. Labeling accuracies were close to 100% for large vessels, but dropped to 50-60% for small vessels that were only present in less than 50% of the set. With this work we showed that using solely spatial information of the vessel labels, vessel segments from stable vessels (>50% presence) were reliably classified. This spatial information will form the basis for a future labeling strategy with a very loose topological model.

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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

DEFF Research Database (Denmark)

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

2012-01-01

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

A study of the external cooling capability for the prevention of reactor vessel failure

Energy Technology Data Exchange (ETDEWEB)

Chang, S H; Baek, W P; Moon, S K; Yang, S H; Kim, S H [Korea Advanced Institute of Science Technology, Daejeon (Korea, Republic of)

1994-07-15

This study (a 3-year program) aims to perform a comprehensive assessment of the feasibility of external vessel flooding with respect to advanced pressurized water reactor plants to be built in Korea. During the first year, review of the relevant phenomena and preliminary assessment of the concept have been performed. Also performed is a review of heat transfer correlations for the computer program that will be developed for assessment of the cooling capability of external vessel flooding. Important phenomena that determine the cooling capability of external vessel flooding are (a) the initial transient before formation of molten corium pool, (b) natural convection of in-vessel molten corium pool, (c) radiative heat exchange between the molten corium pool and the upper vessel structures, (d) thermal hydraulics outside the vessel, (e) structural integrity consideration, and (f) long-term phenomena. The adoption of the concept should be decided by considering several factors such as (a) vessel submergence procedure, (b) cooling requirements, (c) vessel design features, (d) steam production, (e) instrumentation needs, and (f) an overall accident management strategy. The external vessel cooling concept looks to be promising. However, further study is required for a reliable decision making. Several correlations are available for the prediction of cooling capability of the present concept. However, it is difficult to define a sufficiently reliable set of correlations; sensitivity studies would be required in assessing the cooling capability with the computer program.

Neutronics studies for the design of the European DEMO vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Flammini, Davide, E-mail: davide.flammini@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Villari, Rosaria; Moro, Fabio; Pizzuto, Aldo [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Bachmann, Christian [EUROfusion Consortium, Boltzmannstr. 2, 85748 Garching (Germany)

2016-11-01

Highlights: • MCNP calculation of nuclear heating, damage, helium production and neutron flux in DEMO HCLL and HCPB vacuum vessel at the inboard equatorial plane. • Study of impact of the poloidal gap between blanket modules, for several gap width, on vacuum vessel nuclear quantities. • Effect of the gap on nuclear heating result to be moderate, however high values of nuclear heating are found, even far from the gap with HCLL blanket. • Radiation damage limit of 2.75 DPA is met with a 1 cm wide gap. Helium production results very sensitive to the gap width. • Comparison between HCLL and HCPB blankets is shown for nuclear heating and neutron flux in the vacuum vessel. - Abstract: The DEMO vacuum vessel, a massive water cooled double-walled steel vessel, is located behind breeding blankets and manifolds and it will be subjected to an intense neutron and photon irradiation. Therefore, a proper evaluation of the vessel nuclear heat loads is required to assure adequate cooling and, given the significant lifetime neutron fluence of DEMO, the radiation damage limit of the vessel needs to be carefully controlled. In the present work nuclear heating, radiation damage (DPA), helium production, neutron and photon fluxes have been calculated on the vacuum vessel at the inboard by means of MCNP5 using a 3D Helium Cooled Lithium Lead (HCLL) DEMO model with 1572 MW of fusion power. In particular, the effect of the poloidal gap between the breeding-blanket segments on vacuum vessel nuclear loads has been estimated varying the gap width from 0 to 5 cm. High values of the nuclear heating (≈1 W/cm{sup 3}), which might cause intense thermal stresses, were obtained in inboard equatorial zone. The effect of the poloidal gap on the nuclear heating resulted to be moderate (within 30%). The radiation damage limit of 2.75 DPA on the vessel is almost met with 1 cm of poloidal gap over DEMO lifetime. A comparison with Helium Cooled Pebble Bed blanket is also provided.

Study on mitigation of in-vessel release of fission products in severe accidents of PWR

International Nuclear Information System (INIS)

Huang, G.F.; Tong, L.L.; Li, J.X.; Cao, X.W.

2010-01-01

Research highlights: → In-vessel release of fission products in severe accidents for 600 MW PWR is analyzed. → Mitigation effect of primary feed-and-bleed on in-vessel release is investigated. → Mitigation effect of secondary feed-and-bleed on in-vessel release is studied. → Mitigation effect of ex-vessel cooling on in-vessel release is evaluated. - Abstract: During the severe accidents in a nuclear power plant, large amounts of fission products release with accident progression, including in-vessel and ex-vessel release. Mitigation of fission products release is demanded for alleviating radiological consequence in severe accidents. Mitigation countermeasures to in-vessel release are studied for Chinese 600 MW pressurized water reactor (PWR), including feed-and-bleed in primary circuit, feed-and-bleed in secondary circuit and ex-vessel cooling. SBO, LOFW, SBLOCA and LBLOCA are selected as typical severe accident sequences. Based on the evaluation of in-vessel release with different startup time of countermeasure, and the coupling relationship between thermohydraulics and in-vessel release of fission products, some results are achieved. Feed-and-bleed in primary circuit is an effective countermeasure to mitigate in-vessel release of fission products, and earlier startup time of countermeasure is more feasible. Feed-and-bleed in secondary circuit is also an effective countermeasure to mitigate in-vessel release for most severe accident sequences that can cease core melt progression, e.g. SBO, LOFW and SBLOCA. Ex-vessel cooling has no mitigation effect on in-vessel release owing to inevitable core melt and relocation.

The 4-vessel Sampling Approach to Integrative Studies of Human Placental Physiology In Vivo.

Science.gov (United States)

Holme, Ane M; Holm, Maia B; Roland, Marie C P; Horne, Hildegunn; Michelsen, Trond M; Haugen, Guttorm; Henriksen, Tore

2017-08-02

The human placenta is highly inaccessible for research while still in utero. The current understanding of human placental physiology in vivo is therefore largely based on animal studies, despite the high diversity among species in placental anatomy, hemodynamics and duration of the pregnancy. The vast majority of human placenta studies are ex vivo perfusion studies or in vitro trophoblast studies. Although in vitro studies and animal models are essential, extrapolation of the results from such studies to the human placenta in vivo is uncertain. We aimed to study human placenta physiology in vivo at term, and present a detailed protocol of the method. Exploiting the intraabdominal access to the uterine vein just before the uterine incision during planned cesarean section, we collect blood samples from the incoming and outgoing vessels on the maternal and fetal sides of the placenta. When combining concentration measurements from blood samples with volume blood flow measurements, we are able to quantify placental and fetal uptake and release of any compound. Furthermore, placental tissue samples from the same mother-fetus pairs can provide measurements of transporter density and activity and other aspects of placental functions in vivo. Through this integrative use of the 4-vessel sampling method we are able to test some of the current concepts of placental nutrient transfer and metabolism in vivo, both in normal and pathological pregnancies. Furthermore, this method enables the identification of substances secreted by the placenta to the maternal circulation, which could be an important contribution to the search for biomarkers of placenta dysfunction.

Experimental study on in-vessel debris coolability during severe accident

International Nuclear Information System (INIS)

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

2002-05-01

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

The 'Safari' Technique to Perform Difficult Subintimal Infragenicular Vessels

International Nuclear Information System (INIS)

Gandini, Roberto; Pipitone, Vincenzo; Stefanini, Matteo; Maresca, Luciano; Spinelli, Alessio; Colangelo, Vittorio; Reale, Carlo Andrea; Pampana, Enrico; Simonetti, Giovanni

2007-01-01

The purpose of this study was to describe the efficacy of planned combined subintimal arterial flossing with antegrade-retrograde intervention (SAFARI) to obtain the precise recanalization of the patent portion of a distal runoff vessel in critical limb ischemia (CLI) patients presenting long occlusions involving the popliteal trifurcation. Four patients at risk of limb loss due to long occlusions involving the leg vessel tree and not suitable for a surgical bypass were treated by the subintimal antegrade and retrograde (posterior tibial or anterior tibial artery) approach. The patent portion of the runoff vessel was previously assessed by magnetic resonance angiography (MRA) and directly punctured under Doppler ultrasound (US) guidance. A subintimal channel rendezvous was performed to allow snaring of the guidewires. Subsequently, a balloon dilatation was performed without stent deployment. All patients were successfully recanalized and had complete healing of the limb lesions. At the 12-month follow-up all patients showed clinical improvement with no major complications related to the procedure. This combined antegrade and retrograde subintimal approach is currently an excellent endovascular option in patients with long occlusions extending onto the leg vessels trifurcation and at risk of limb loss

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.

Design study on steam generator integration into the VVER reactor pressure vessel

International Nuclear Information System (INIS)

Hort, J.; Matal, O.

2004-01-01

The primary circuit of VVER (PWR) units is arranged into loops where the heat generated by the reactor is removed by means of main circulating pumps, loop pipelines and steam generators, all located outside the reactor pressure vessel. If the primary circuit and reactor core were integrated into one pressure vessel, as proposed, e.g., within the IRIS project (WEC), a LOCA situation would be limited by the reactor pressure vessel integrity only. The aim of this design study regarding the integration of the steam generator into the reactor pressure vessel was to identify the feasibility limits and some issues. Fuel elements and the reactor pressure vessel as used in the Temelin NPP were considered for the analysis. From among the variants analyzed, the variant with steam generators located above the core and vertically oriented circulating pumps at the RPV lower bottom seems to be very promising for future applications

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

International Nuclear Information System (INIS)

Arendts, J.G.

1982-01-01

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

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

International Nuclear Information System (INIS)

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

1999-04-01

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

77 FR 36998 - Proposed Information Collection; Comment Request; Mandatory Shrimp Vessel and Gear...

Science.gov (United States)

2012-06-20

... Collection; Comment Request; Mandatory Shrimp Vessel and Gear Characterization Survey AGENCY: National... under the Shrimp Fishery Management Plan (FMP). The regulations for the Gulf Shrimp Vessel and Gear... and Gear Characterization Form. The form will be provided by NMFS at the time of permit application...

78 FR 40436 - Proposed Information Collection; Comment Request; Mandatory Shrimp Vessel and Gear...

Science.gov (United States)

2013-07-05

... Collection; Comment Request; Mandatory Shrimp Vessel and Gear Characterization Survey AGENCY: National... under the Shrimp Fishery Management Plan (FMP). The regulations for the Gulf Shrimp Vessel and Gear... Gear Characterization Form. The form will be provided by NMFS at the time of permit application and...

Effect of radiation damage on operating safety of steel pressure vessels of nuclear reactors

International Nuclear Information System (INIS)

Vacek, M.; Havel, S.; Stoces, B.; Brumovsky, M.

1980-01-01

The effects are assessed of the environment upon mechanical properties of steel used generally for pressure vessels of light water nuclear reactors. Changes caused by radiation affect the reliability of vessels. Deterioration of steel properties is mainly due to neutron radiation. The article deals with factors bearing upon damage and with methods allowing to evaluate the reliability of vessels and predict their service life. Operating reliability of vessels is very unfavourably affected by planned and accidental reactor transients. (author)

Evaluation Plan on In-vessel Source Term in PGSFR

Energy Technology Data Exchange (ETDEWEB)

Lee, Seung Won; Ha, Kwi-Seok; Ahn, Sang June; Lee, Kwi Lim; Jeong, Taekyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

This strategy requires nuclear plants to have features that prevent radionuclide release and multiple barriers to the escape from the plants of any radionuclides that are released despite preventive measures. Considerations of the ability to prevent and mitigate release of radionuclides arise at numerous places in the safety regulations of nuclear plants. The effectiveness of mitigative capabilities in nuclear plants is subject to quantitative analysis. The radionuclide input to these quantitative analyses of effectiveness is the Source Term (ST). All features of the composition, magnitude, timing, chemical form and physical form of accidental radionuclide release constitute the ST. Also, ST is defined as the release of radionuclides from the fuel and coolant into the containment, and subsequently to the environment. The in-vessel STs of PGSFR will be estimated using the methodology of ANL-ART-38 report in additional to 4S methodology. The in-vessel STs are calculated through several phases: The inventory of each radionuclide is calculated by ORIGEN-2 code using the realistic burnup conditions. ST in the release from the core to primary sodium is calculated by using the assumption of ANL methodology. Lastly, ST in the release from the primary sodium to cover gas space is calculated by using equation and experimental materials.

Evaluation Plan on In-vessel Source Term in PGSFR

International Nuclear Information System (INIS)

Lee, Seung Won; Ha, Kwi-Seok; Ahn, Sang June; Lee, Kwi Lim; Jeong, Taekyeong

2016-01-01

This strategy requires nuclear plants to have features that prevent radionuclide release and multiple barriers to the escape from the plants of any radionuclides that are released despite preventive measures. Considerations of the ability to prevent and mitigate release of radionuclides arise at numerous places in the safety regulations of nuclear plants. The effectiveness of mitigative capabilities in nuclear plants is subject to quantitative analysis. The radionuclide input to these quantitative analyses of effectiveness is the Source Term (ST). All features of the composition, magnitude, timing, chemical form and physical form of accidental radionuclide release constitute the ST. Also, ST is defined as the release of radionuclides from the fuel and coolant into the containment, and subsequently to the environment. The in-vessel STs of PGSFR will be estimated using the methodology of ANL-ART-38 report in additional to 4S methodology. The in-vessel STs are calculated through several phases: The inventory of each radionuclide is calculated by ORIGEN-2 code using the realistic burnup conditions. ST in the release from the core to primary sodium is calculated by using the assumption of ANL methodology. Lastly, ST in the release from the primary sodium to cover gas space is calculated by using equation and experimental materials

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-01

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

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)

Reactor vessel nozzle cracks: a photoelastic study

International Nuclear Information System (INIS)

Smith, C.W.

1979-01-01

A method consisting of a marriage between the ''frozen stress'' photoelastic approach and the local stress field equations of linear elastic fracture mechanics for estimating stress intensity factor distributions in three dimensional, finite cracked body problems is reviewed and extensions of the method are indicated. The method is then applied to the nuclear reactor vessel nozzle corner crack problem for both Intermediate Test Vessel and Boiling Water Reactor geometries. Results are compared with those of other investigators. 35 refs

USCG Security Plan Review

Data.gov (United States)

Department of Homeland Security — The Security Plan Review module is intended for vessel and facility operators to check on the status of their security plans submitted to the US Coast Guard. A MISLE...

Simulative study on dose distribution of 103Pd stent in blood-vessel

International Nuclear Information System (INIS)

Yuan Shuyu; Dai Guangfu; Xu Zhiyong; Sun Fuyin; Xu Shuhe; Ma Fengwu

2003-01-01

Objective: To evaluate the dose distribution of 103 Pd stent in the blood-vessel. Methods: Simulative study on dose distribution of endovascular 103 Pd stent was conducted with thermoluminescence dosimeter. The vessel wall was substituted by muscle equivalent material in this simulative study. Results: When radioactivity of the study 103 Pd stent was 9.8 MBq the absorbed dose from the stent surface by muscle equivalent material was 9.8 Gy at 17 d (the half-life period of 103 Pd). The radioactivity of 103 Pd stent surface rapidly attenuated over the radial distance. 80% of the radioactivity at the area that was radially 0.4 mm apart from the stent surface was absorbed by the simulative blood-vessel wall. Conclusion: Endovascular 103 Pd stent does not exert significant injury on the surrounding organs or tissues

Scoping Study of Airlift Circulation Technologies for Supplemental Mixing in Pulse Jet Mixed Vessels

Energy Technology Data Exchange (ETDEWEB)

Schonewill, Philip P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Berglin, Eric J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Boeringa, Gregory K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buchmiller, William C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burns, Carolyn A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Minette, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-04-07

At the request of the U.S. Department of Energy Office of River Protection, Pacific Northwest National Laboratory (PNNL) conducted a scoping study to investigate supplemental technologies for supplying vertical fluid motion and enhanced mixing in Waste Treatment and Immobilization Plant (WTP) vessels designed for high solids processing. The study assumed that the pulse jet mixers adequately mix and shear the bottom portion of a vessel. Given that, the primary function of a supplemental technology should be to provide mixing and shearing in the upper region of a vessel. The objective of the study was to recommend a mixing technology and configuration that could be implemented in the 8-ft test vessel located at Mid-Columbia Engineering (MCE). Several mixing technologies, primarily airlift circulator (ALC) systems, were evaluated in the study. This technical report contains a review of ALC technologies, a description of the PNNL testing and accompanying results, and recommended features of an ALC system for further study.

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)

In-vessel core debris retention experiments. Final report

International Nuclear Information System (INIS)

1998-10-01

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

An experimental system for the study of ultrasound exposure of isolated blood vessels

NARCIS (Netherlands)

Tokarczyk, Anna; Rivens, Ian; van Bavel, E.; Symonds-Tayler, Richard; ter Haar, Gail

2013-01-01

An experimental system designed for the study of the effects of diagnostic or therapeutic ultrasound exposure on isolated blood vessels in the presence or absence of intraluminal contrast agent is described. The system comprised several components. A microscope was used to monitor vessel size (and

Surveillance specimen programmes for WWER reactor vessels in the Czech Republic

International Nuclear Information System (INIS)

Brynda, J.; Hogel, J.; Brumovsky, M.

2003-01-01

The present state of materials degradation in WWER reactor pressure vessels manufactured in the Czech Republic is highlighted. The standard surveillance program for WWER-440/V-213 type reactors is described and its deficiencies together with the main results obtained are discussed. A new supplementary surveillance program meeting all requirements for PWR type reactors has been developed and launched. An entirely new design was chosen for the surveillance programme for WWER-1000/V-320 type reactor pressure vessels. Materials selection, container design and location as well as the withdrawal plan connected with ex-vessel fluence monitoring are described

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)

Lessons learned from the NRU vessel leak repair and return to service projects

International Nuclear Information System (INIS)

Heeney, P.; Turcotte, J.

2011-01-01

In May 2009 the National Research Universal (NRU) reactor was shut down due to a small leak detected from the reactor vessel into the annulus surrounding the reactor. What ensued was a challenging, yet successful, 15 month long Repair and Return to Service Outage. This Repair and Return to Service Outage presented many first-of-a-kind challenges that provide learning opportunities which have been incorporated into subsequent planned outages. These lessons learned are invaluable tools to be used in the planning and execution of future outages. Following the repair of the NRU vessel, AECL was required to conduct annual inspections of the vessel wall. These inspections require an annual Extended Outage (up to 4 weeks in length). A planned Extended Outage was conducted in May/June 2011 and provided an opportunity to implement some of the lessons learned during the Repair and Return to Service Outage. Lessons learned from that Extended Outage have been incorporated in the subsequent monthly maintenance outages, with lessons learned sessions being held after each outage to ensure that the execution of outages is constantly improving. (author)

Instrumentation and testing of a prestressed concrete containment vessel model

International Nuclear Information System (INIS)

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

1997-01-01

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

Macular Choroidal Small-Vessel Layer, Sattler's Layer and Haller's Layer Thicknesses: The Beijing Eye Study.

Science.gov (United States)

Zhao, Jing; Wang, Ya Xing; Zhang, Qi; Wei, Wen Bin; Xu, Liang; Jonas, Jost B

2018-03-13

To study macular choroidal layer thickness, 3187 study participants from the population-based Beijing Eye Study underwent spectral-domain optical coherence tomography with enhanced depth imaging for thickness measurements of the macular small-vessel layer, including the choriocapillaris, medium-sized choroidal vessel layer (Sattler's layer) and large choroidal vessel layer (Haller's layer). In multivariate analysis, greater thickness of all three choroidal layers was associated (all P  0.05) associated with the prevalence of open-angle glaucoma or diabetic retinopathy. There was a tendency (0.07 > P > 0.02) toward thinner choroidal layers in chronic angle-closure glaucoma. The ratio of small-vessel layer thickness to total choroidal thickness increased (P layer and Haller's layer thickness to total choroidal thickness decreased. A higher ratio of small-vessel layer thickness to total choroidal thickness was significantly associated with a lower prevalence of AMD (early type, intermediate type, late geographic type). Axial elongation-associated and aging-associated choroidal thinning affected Haller's and Sattler's layers more markedly than the small-vessel layer. Non-exudative and exudative AMD, except for geographic atrophy, was associated with slightly increased choroidal thickness.

Preliminary study of an expert system for mechanical design of a pressure vessel

International Nuclear Information System (INIS)

Kasmuri, N.H.; Md Som, A.

2006-01-01

This paper describes a preliminary study of an expert system for mechanical design of a pressure vessel. The system supports the framework for the conceptual mechanical design from the initial stages within the design procedures. ASME Boiler and Pressure Vessel Code Section VIII Division 1 were applied as a design rule. The proposed methodology facilitates the development of knowledge base acquisition, knowledge base construction and the prototype implementation. This study characterizes a knowledge base (procedure) of mechanical design of a pressure vessel subjected to internal pressure including all design parameters; i.e. temperature, shell thickness, selection of materials of constructions, stress analysis procedure, support and ancillary items. The rationalization of the mechanical design is shown in the form of a schematic flow diagram. A Kappa PC expert system shell is used as a tool to develop the prototype software. It provides graphical representation for creating objects, hierarchies and rules for knowledge base used in pressure vessel design. (Author)

Conceptual design studies of in-vessel viewing equipment for ITER

International Nuclear Information System (INIS)

Obara, Kenjiro; Oka, Kiyoshi; Taguchi, Hiroshi; Itoh, Akira; Tada, Eisuke; Shibanuma, Kiyoshi

1996-03-01

In-vessel viewing systems are essential to inspect all surface of in-vessel components so as to detect and locate damages, and to assist in-vessel maintenance operations. The in-vessel viewing operations are categorized into the three cases, which are 1) rapid inspection just after off-normal events such as disruption, 2) scheduled inspection, and 3) supplementary inspection during maintenance operations. In case of the rapid inspection, the viewing systems have to be operated in vacuum (ca. 10 -5 Pa) and high temperature (ca. 300degC) under a gamma ray dose rate of 10 7 R/h. On the other hand, the latter two cases are anticipated to be under atmospheric inert gas, 150degC and 3x10 6 R/h. Accordingly, the in-vessel viewing systems are required to have sufficient durability under those conditions of all cases as well as precision of the vision to all of in-vessel surface. Based on those requirements, scoping studies on various viewing concepts have been performed and the applicability to the ITER conditions have been assessed. As a result, two types of viewing systems have been chosen, which are a periscope type viewing system and a image fiber type viewing system with a multi-joint manipulator. Both systems are based on radiation hard optical elements which are being developed. In this report, the design features of both viewing systems are described, including technical issues for ITER application. Finally, a periscope type viewing system is recommended as a primary system and the following specifications/conditions are proposed for the further engineering design. (1) Unified type periscope with a movable mirror at the tip (2) Integrated lighting device into the periscope (3) Accessed from top vertical ports located at 7.3m from the machine center (4) Proposed configuration with a total length of around 27m and a diameter of 200mm. (author)

Generating Optimal Stowage Plans for Container Vessel Bays

DEFF Research Database (Denmark)

Delgado-Ortegon, Alberto; Jensen, Rune Møller; Schulte, Christian

2009-01-01

collaboration to date with a liner shipping company on automated stowage planning. We then show how to solve this model ef- ficiently in - to our knowledge - the first application of CP to stowage planning using state-of-the-art techniques such as extensive use of global constraints, viewpoints, static...

Best practices for preparing vessel internals segmentation projects

International Nuclear Information System (INIS)

Boucau, Joseph; Segerud, Per; Sanchez, Moises

2016-01-01

Westinghouse has been involved in reactor internals segmentation activities in the U.S. and Europe for 30 years. Westinghouse completed in 2015 the segmentation of the reactor vessel and reactor vessel internals at the Jose Cabrera nuclear power plant in Spain and a similar project is on-going at Chooz A in France. For all reactor dismantling projects, it is essential that all activities are thoroughly planned and discussed up-front together with the customer. Detailed planning is crucial for achieving a successful project. One key activity in the preparation phase is the 'Segmentation and Packaging Plan' that documents the sequential steps required to segment, separate, and package each individual component, based on an activation analysis and component characterization study. Detailed procedures and specialized rigging equipment have to be developed to provide safeguards for preventing certain identified risks. The preparatory work can include some plant civil structure modifications for making the segmentation work easier and safer. Some original plant equipment is sometimes not suitable enough and need to be replaced. Before going to the site, testing and qualification are performed on full scale mock-ups in a specially designed pool for segmentation purposes. The mockup testing is an important step in order to verify the function of the equipment and minimize risk on site. This paper is describing the typical activities needed for preparing the reactor internals segmentation activities using under water mechanical cutting techniques. It provides experiences and lessons learned that Westinghouse has collected from its recent projects and that will be applied for the new awarded projects. (authors)

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)

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

International Nuclear Information System (INIS)

Jin Di; Liu Xiaojing; Cheng Xu; Li Fei

2014-01-01

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

TMI-2 reactor vessel head removal

International Nuclear Information System (INIS)

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

1984-12-01

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

TMI-2 reactor vessel head removal

International Nuclear Information System (INIS)

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

1985-09-01

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

Status of pressure vessel embrittlement study in Japan

Energy Technology Data Exchange (ETDEWEB)

Kataoka, Shigeki [Japan Power Engineering and Inspection Corp. (JAPEIC), Chiba (Japan)

1997-09-01

The number of nuclear power plants in service for more than 20 years is increasing in Japan. Subsequently, the aging of nuclear power plants will continue to increase and for this reason, the assurance of the safety and reliability of nuclear power plants is becoming more important. Under this circumstances, Japan Government issued a report: ``Specific Concepts in Dealing with Nuclear Power Plant High Aging`` in April, 1996. This report identified that continuous technology development efforts are important to deal with the issues of nuclear power plant aging, and the following items are extracted for important categories to be developed. (1) Aging phenomena evaluation technology. (2) Inspection/monitoring technology (3) Preventive maintenance/repair technology. Japan Power Engineering and Inspection Corporation (JAPEIC) have been implementing various verification test concerning the above items consigned by the Ministry of International Trade and Industry (MITI). This report outlines the Specific Concepts in Dealing with Nuclear Power Plant High Agency and the past achievements and future plans of various verification tests related to irradiation embrittlement of nuclear reactor pressure vessel, mainly related to Pressurized Thermal Shock (PTS). (author). 4 refs, 8 figs, 5 tabs.

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

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

International Nuclear Information System (INIS)

Whitman, G.D.

1975-01-01

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

Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

Energy Technology Data Exchange (ETDEWEB)

Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

2001-03-15

In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report.

Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

International Nuclear Information System (INIS)

Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin; Park, Jae Hong

2001-03-01

In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report

Path Planning Algorithms for the Adaptive Sensor Fleet

Science.gov (United States)

Stoneking, Eric; Hosler, Jeff

2005-01-01

The Adaptive Sensor Fleet (ASF) is a general purpose fleet management and planning system being developed by NASA in coordination with NOAA. The current mission of ASF is to provide the capability for autonomous cooperative survey and sampling of dynamic oceanographic phenomena such as current systems and algae blooms. Each ASF vessel is a software model that represents a real world platform that carries a variety of sensors. The OASIS platform will provide the first physical vessel, outfitted with the systems and payloads necessary to execute the oceanographic observations described in this paper. The ASF architecture is being designed for extensibility to accommodate heterogenous fleet elements, and is not limited to using the OASIS platform to acquire data. This paper describes the path planning algorithms developed for the acquisition phase of a typical ASF task. Given a polygonal target region to be surveyed, the region is subdivided according to the number of vessels in the fleet. The subdivision algorithm seeks a solution in which all subregions have equal area and minimum mean radius. Once the subregions are defined, a dynamic programming method is used to find a minimum-time path for each vessel from its initial position to its assigned region. This path plan includes the effects of water currents as well as avoidance of known obstacles. A fleet-level planning algorithm then shuffles the individual vessel assignments to find the overall solution which puts all vessels in their assigned regions in the minimum time. This shuffle algorithm may be described as a process of elimination on the sorted list of permutations of a cost matrix. All these path planning algorithms are facilitated by discretizing the region of interest onto a hexagonal tiling.

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.

Results of steel containment vessel model test

International Nuclear Information System (INIS)

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

1998-05-01

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

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

International Nuclear Information System (INIS)

Ibe, Yukimi; Kitajima, Masatake

1977-01-01

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

Feasibility Study on Two-phase Thermosiphon for External Vessel Cooling Application of SFR

Energy Technology Data Exchange (ETDEWEB)

Choi, Jae Young; Song, Sub Lee; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-05-15

This study shows that ex-vessel cooling by two-phase thermosiphon is feasible for large size of SFR. The result presents that further studies to increase heat transfer on condenser-air and gap is necessary and the experiment should be conducted for the validation. Also, the heat loss through evaporator during normal operation, corrosion, consideration of organic fluid to exclude the poison of mercury should be studied. As the necessity of sodium fast reactor in order to reduce spent fuel, the development of designing sodium fast reactor becomes an issue. Even though there is PDRC and RVACS for the decay heat removal (DHR) system, each system has disadvantage of sodium fire and low performance, respectively. Therefore, to increase the safety of SFR, the new passive safety system design is needed without sodium fire and high performance, which can applied for large SFR. The DHR system using two-phase thermosiphon for external vessel cooling application is suggested in this paper. The proposed design have advantage that there is no structure in reactor vessel, which means no system modification and no sodium fire with perfect isolation. Also, it provide the method to mitigate sodium fire in case of sodium leakage from reactor vessel.

Progressive Macula Vessel Density Loss in Primary Open-Angle Glaucoma: A Longitudinal Study.

Science.gov (United States)

Shoji, Takuhei; Zangwill, Linda M; Akagi, Tadamichi; Saunders, Luke J; Yarmohammadi, Adeleh; Manalastas, Patricia Isabel C; Penteado, Rafaella C; Weinreb, Robert N

2017-10-01

To characterize the rate of macula vessel density loss in primary open-angle glaucoma (POAG), glaucoma-suspect, and healthy eyes. Longitudinal, observational cohort from the Diagnostic Innovations in Glaucoma Study. One hundred eyes (32 POAG, 30 glaucoma-suspect, and 38 healthy) followed for at least 1 year with optical coherence tomography angiography (OCT-A) imaging on at least 2 visits were included. Vessel density was calculated in the macula superficial layer. The rate of change was compared across diagnostic groups using a multivariate linear mixed-effects model. Baseline macula vessel density was highest in healthy eyes, followed by glaucoma-suspect and POAG eyes (P macula whole en face vessel density was significantly faster in glaucoma eyes (-2.23%/y) than in glaucoma-suspect (0.87%/y, P = .001) or healthy eyes (0.29%/y, P = .004). Conversely, the rate of change in ganglion cell complex (GCC) thickness was not significantly different from zero in any diagnostic group, and no significant differences in the rate of GCC change among diagnostic groups were found. With a mean follow-up of less than 14 months, eyes with POAG had significantly faster loss of macula vessel density than either glaucoma-suspect or healthy eyes. Serial OCT-A measurements also detected glaucomatous change in macula vessel density in eyes without evidence of change in GCC thickness. Copyright © 2017 Elsevier Inc. All rights reserved.

Immersive volume rendering of blood vessels

Science.gov (United States)

Long, Gregory; Kim, Han Suk; Marsden, Alison; Bazilevs, Yuri; Schulze, Jürgen P.

2012-03-01

In this paper, we present a novel method of visualizing flow in blood vessels. Our approach reads unstructured tetrahedral data, resamples it, and uses slice based 3D texture volume rendering. Due to the sparse structure of blood vessels, we utilize an octree to efficiently store the resampled data by discarding empty regions of the volume. We use animation to convey time series data, wireframe surface to give structure, and utilize the StarCAVE, a 3D virtual reality environment, to add a fully immersive element to the visualization. Our tool has great value in interdisciplinary work, helping scientists collaborate with clinicians, by improving the understanding of blood flow simulations. Full immersion in the flow field allows for a more intuitive understanding of the flow phenomena, and can be a great help to medical experts for treatment planning.

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Distributed situation awareness in complex collaborative systems: A field study of bridge operations on platform supply vessels.

Science.gov (United States)

Sandhåland, Hilde; Oltedal, Helle A; Hystad, Sigurd W; Eid, Jarle

2015-06-01

This study provides empirical data about shipboard practices in bridge operations on board a selection of platform supply vessels (PSVs). Using the theoretical concept of distributed situation awareness, the study examines how situation awareness (SA)-related information is distributed and coordinated at the bridge. This study thus favours a systems approach to studying SA, viewing it not as a phenomenon that solely happens in each individual's mind but rather as something that happens between individuals and the tools that they use in a collaborative system. Thus, this study adds to our understanding of SA as a distributed phenomenon. Data were collected in four field studies that lasted between 8 and 14 days on PSVs that operate on the Norwegian continental shelf and UK continental shelf. The study revealed pronounced variations in shipboard practices regarding how the bridge team attended to operational planning, communication procedures, and distracting/interrupting factors during operations. These findings shed new light on how SA might decrease in bridge teams during platform supply operations. The findings from this study emphasize the need to assess and establish shipboard practices that support the bridge teams' SA needs in day-to-day operations. Provides insights into how shipboard practices that are relevant to planning, communication and the occurrence of distracting/interrupting factors are realized in bridge operations.Notes possible areas for improvement to enhance distributed SA in bridge operations.

Computational study of the mixed cooling effects on the in-vessel retention of a molten pool in a nuclear reactor

International Nuclear Information System (INIS)

Kim, Byung Seok; Sohn, Chang Hyun; Ahn, Kwang Il

2004-01-01

The retention of a molten pool vessel cooled by internal vessel reflooding and/or external vessel reactor cavity flooding has been considered as one of severe accident management strategies. The present numerical study investigates the effect of both internal and external vessel mixed cooling on an internally heated molten pool. The molten pool is confined in a hemispherical vessel with reference to the thermal behavior of the vessel wall. In this study, our numerical model used a scaled-down reactor vessel of a KSNP (Korea Standard Nuclear Power) reactor design of 1000 MWe (a pressurized water reactor with a large and dry containment). Well-known temperature-dependent boiling heat transfer curves are applied to the internal and external vessel cooling boundaries. Radiative heat transfer has been considered in the case of dry internal vessel boundary condition. Computational results show that the external cooling vessel boundary conditions have better effectiveness than internal vessel cooling in the retention of the melt pool vessel failure

A GRASP algorithm for the container stowage slot planning problem

DEFF Research Database (Denmark)

Parreno, Francisco; Pacino, Dario; Alvarez-Valdes, Ramon

2016-01-01

in clusters along the vessel. For each of those clusters a specific position for each container must be found. Compared to previous studies, we have introduced two new features: the explicit handling of rolled out containers and the inclusion of separations rules for dangerous cargo. We present a novel......This work presents a generalization of the Slot Planning Problem which raises when the liner shipping industry needs to plan the placement of containers within a vessel (stowage planning). State-of-the-art stowage planning relies on a heuristic decomposition where containers are first distributed...... integer programming formulation and a Greedy Randomized Adaptive Search Procedure (GRASP) to solve the problem. The approach is able to find high-quality solution within 1 s. We also provide comparison with the state-of-the-art on an existing and a new set of benchmark instances. (C) 2016 Elsevier Ltd...

DLC coating of textile blood vessels using PLD

Czech Academy of Sciences Publication Activity Database

Kocourek, Tomáš; Jelínek, Miroslav; Vorlíček, Vladimír; Zemek, Josef; Janča, T.; Žížková, V.; Podlaha, J.; Popov, C.

2008-01-01

Roč. 93, č. 3 (2008), s. 627-632 ISSN 0947-8396 R&D Projects: GA MPO FI-IM2/068; GA ČR GA202/06/0216 Institutional research plan: CEZ:AV0Z10100522 Keywords : blood vessels * PLD * DLC * sp2 * sp3 Subject RIV: BH - Optics, Masers, Lasers Impact factor: 1.884, year: 2008

46 CFR 122.503 - Voyage plan.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Voyage plan. 122.503 Section 122.503 Shipping COAST... Emergencies § 122.503 Voyage plan. (a) The master of the following vessels shall prepare a voyage plan: (1) A... United States Great Lakes port from a Canadian Great Lakes port. (b) The voyage plan required by...

75 FR 54026 - Salvage and Marine Firefighting Requirements; Vessel Response Plans for Oil

Science.gov (United States)

2010-09-03

... information for resource providers for each vessel with appropriate equipment and resources located in each... appropriate salvage and marine firefighting resources were identified and available for responding to...

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.

Examination of VVER-1000 Reactor Pressure Vessel

International Nuclear Information System (INIS)

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

2008-01-01

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

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

In vessel core melt progression phenomena

International Nuclear Information System (INIS)

Courtaud, M.

1993-01-01

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

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).

Performance of evaluation techniques for planning preventive repair of leakages on BWR vessel stub tubes

International Nuclear Information System (INIS)

Prat, J.P.

1985-01-01

Stub tubes have been installed on the control rod drive assemblies of the majority of boiling water reactor vessel bottoms. These stub tubes can undergo a slow process of cracking, producing no potential hazard but occasional plant unavailability due to leakage in excess of specification limits. In the case considered here, the development and application of remotely controlled equipment permitting access both from inside and outside the housing, and of crack evolution and location techniques, has made the progressive development, selection and planning of repairs a possibility, thus minimizing the risk of leakages appearing between refuelling outages. The example used here is of a Nuclear Power Plant in which a first leak appeared in 1981. Since that time extensive and progressive laboratory and field development work has been carried out. This paper presents the experience gained with customized, remote-control inspection equipment, and several developed or adapted techniques, as the basis for an in-service repair decision tree, the case in question being that of a BWR plant in which a cracking process, not affecting safety but potentially affecting availability, has given rise to leaks after ten years of operation

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

MFTF-α+T end cell vacuum vessel and nuclear shield trade studies

International Nuclear Information System (INIS)

Kirchner, J.

1984-01-01

Three separate and distinct vacuum vessel and nuclear shield trade studies were performed in series. The studies are: vacuum topology, nuclear shield location and composition, and water bulk shield location and material selection

Defense Utility of Commercial Vessels and Craft.

Science.gov (United States)

1980-01-01

interpreted to determine whether the vessel is suitable for the salvage * job required. Sometimes, a photograph or general arrangement plan is most...30. UNDERWAY REPLENISHMENT (yes/no) Yes 31 . OTHER POSSIBLE FEATURES Diesel powered mue pur’ps 2000-4000 HP, Draw works 2000-3000 HTP C. BARGE...of the Army, Marine Crewman’s Handbook, Technical Manual TM 55-501 (Washington: Sept. 1972). 23. Department of the Army, Marine Equipment

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.

Turkish Republic of Northern Cyprus Vessel Traffic Services (TRNCVTS

Directory of Open Access Journals (Sweden)

Serdar KUM

2016-06-01

Full Text Available The first Vessel Traffic Service (VTS started in 1949 in the Liverpool Port (UK and it continued in Netherlands in 1956. In Turkey, planning and management of the marine traffic using the waterways and ports in Turkey started with Turkish Straits VTS which came into service in 2003 due to the increase in traffic density enhance its effectiveness and necessity every year. Feasibility studies in five new areas have been initiated for the establishment of the VTS system by force of the strategic decision taken by the maritime authority in 2008. These areas are; İzmit Bay, Izmir Bay and Aliağa Region, Gulf of İskenderun and Mersin. Monitoring the marine traffic has an important place as the Turkish Republic of Northern Cyprus (TRNC is an important transition point in the Eastern Mediterranean region. For this reason, in this study the impact assessment and necessity of the establishment of a VTS to be located in the TRNC were evaluated by using Environmental (PEST: Political, Economic, Social, Technological and SWOT (Strengths, Weakness, Opportunities, Threats Analyses. In addition, the suitability of the possible locations of Traffic Monitoring Stations (TMS has been examined by the field study. Evaluation of TRNC VTS in geographical and strategic terms will benefit from the opportunities and facilities that both the Republic of Turkey and the TRNC will have as a part of the effect of VTS against the embargo imposed on the TRNC in the world. Three appropriate VTS TMSs have been chosen as a result of the field study carried out for 10 determined TMSs; Cape Kormakitis, Cape Apostolos Andreas and Famagusta. It is also concluded that it would be appropriate to plan the Famagusta Station as VTS Centre. It is assumed to monitor and track the vessels in the zones out of the coverage area of these stations by Automatic Identification System (AIS. Safety and security in the shipping, protection of navigation, life, property and the marine environment of

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

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

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

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

Generic BWR-4 degraded core in-vessel study. Status report

International Nuclear Information System (INIS)

1984-11-01

Original intent of this project was to produce a phenomenological study of the in-vessel degradation which occurs during the TQUX and TQUV sequences for a generic BWR-4 from the initiation of the FSAR Chapter 15 operational transient through core debris bed formation to the failure of the primary pressure boundary. Bounding calculations were to be performed for the two high pressure and low pressure non-LOCA scenarios to assess the uncertainties in the current state of knowledge regarding the source terms for containment integrity studies. Source terms as such were defined in terms of hydrogen generation, unreacted metal, and coolant inventroy, and in terms of the form, sequencing and mode of dispersal through the primary vessel boundary. Fission product release was not to be considered as part of this study. Premature termination of the project, however, led to the dicontinuation of work on an as is basis. Work on the in-core phase from the point of scram to core debris bed formation was largely completed. A preliminary scoping calculation on the debris bed phase had been initiated. This report documents the status of the study at termination

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...

Numerical effects in the neutron flux calculations into WWER-type reactor vessels by Monte Carlo method

International Nuclear Information System (INIS)

Alvarez Cardona, C.M.; Rodriguez Gual, M.; Hernandez Valle, S.

2001-01-01

The calculation of neutron fluxes and fluence into reactor pressure vessel is a regulatory requirement in the stages of the design, operation and plan lifetime extension. The reactor vessel is considered a unique and non-substitutable part of the NPP that undergoes degradation. The main source of the aging comes from the fast neutron damage induced in the steel crystalline lattice. Due to the proximity of the core edge to the vessel inner surface; the vessel steel is exposed to high fast neutron fluence. The effect of this irradiation on the mechanical properties becomes more acute because of the impurities measured in the Russian steel alloys. In the present paper, a PC version of the Monte Carlo 3-D HEXANN-EVALU system is used for the estimation of the WWER reactor pressure vessel irradiation. It was selected on the basis of its flexible options that on the other hand need to be quantified in connection with the desired magnitudes. The parameters that control the random walk of neutrons as well as the efficiency increasing options included in the code are studied in order to identify their impact in the final results for fluxes and fluence in the reactor pressure vessel. As a result an optimal set of parameters is suggested. (authors)

Technology and Organisation of Inka Pottery Production in the Leche Valley. Part II: Study of Fired Vessels

International Nuclear Information System (INIS)

Hayashida, F.; Haeusler, W.; Riederer, J.; Wagner, U.

2003-01-01

Ceramic finds from the Inka workshops at Tambo Real and La Vina in the Leche Valley in northern Peru were studied by Moessbauer spectroscopy, thin section microscopy and X-ray diffraction. Sherds of Inka style vessels and of local style vessels can be distinguished by their shape, although local techniques appear to have been used in making both types. A reconstruction of the firing techniques by scientific studies of the ceramic material does not reveal a substantial difference in material or in the firing of both forms, although high firing temperatures were necessary to achieve sufficient stability of the large Inka style vessels. It cannot be decided whether the smaller local vessels were fired together with the Inka vessels or separately. Most of the variation in the maximum firing temperature can be explained with the normal temperature and atmosphere fluctuations in an open pit kiln.

Vessel encoded arterial spin labeling with cerebral perfusion: preliminary study

International Nuclear Information System (INIS)

Wu Bing; Xiao Jiangxi; Xie Cheng; Wang Xiaoying; Jiang Xuexiang; Wong, E.C.; Wang Jing; Guo Jia; Zhang Beiru; Zhang Jue; Fang Jing

2008-01-01

Objective: To evaluate a noninvasive vessel encoded imaging for selective mapping of the flow territories of the left and fight internal carotid arteries and vertebral-basilar arteries. Methods: Seven volunteers [(33.5 ± 4.1) years; 3 men, 4 women] and 6 patients [(55.2 ± 3.2) years; 2 men, 4 women] were given written informed consent approved by the institutional review board before participating in the study. A pseudo-continuous tagging pulse train is modified to encode all vessels of interest. The selectivity of this method was demonstrated. Regional perfusion imaging was developed on the same arterial spin labeling sequence. Perfusion-weighted images of the selectively labeled cerebral arteries were obtained by subtraction of the labeled from control images. The CBF values of hemisphere, white matter, and gray matter of volunteers were calculated. The vessel territories on patients were compared with DSA. The low perfusion areas were compared with high signal areas on T 2 -FLAIR. Results: High SNR maps of left carotid, right carotid, and basilar territories were generated in 8 minutes of scan time. Cerebral blood flow values measured with regional perfusion imaging in the complete hemisphere (32.6 ± 4.3) ml·min -1 · 100 g -1 , white matter (10.8 ± 0.9) ml·min -1 ·100 g -1 , and gray matter (55.6±2.9) ml·min -1 · 100 g -1 were in agreement with data in the literature. Vessel encoded imaging in patients had a good agreement with DSA. The low perfusion areas were larger than high signal areas on T 2 -FLAIR. Conclusion: We present a new method capable of evaluating both quantitatively and qualitatively the individual brain- feeding arteries in vivo. (authors)

Thermal Behavior of the Reactor Vessel Penetration Under External Vessel Cooling During a Severe Accident

International Nuclear Information System (INIS)

Kang, Kyoung-Ho; Park, Rae-Joon; Kim, Jong-Tae; Min, Byung-Tae; Lee, Ki-Young; Kim, Sang-Baik

2004-01-01

Experimental and analytical studies on the thermal behavior of reactor vessel penetration have been performed under external vessel cooling during a severe accident in the Korean next-generation reactor APR1400. Two types of tests, SUS-EXT and SUS-DRY with and without external vessel cooling, respectively, have been performed using sustained heating by an induction heater. Three tests have been carried out varying the cooling conditions at the vessel outer surface in the SUS-EXT tests. The experimental results have been thermally estimated using the LILAC computer code. The experimental results indicate that the inner surface of the vessel was ablated by the 45-mm thickness in the SUS-DRY test. Despite the total ablation of the welding material, the penetration was not ejected outside the vessel, which could be attributed to the thermal expansion of the penetration. Unlike the SUS-DRY test, the thickness of the ablation was ∼15 to 20 mm at most, so the welding was preserved in the SUS-EXT tests. It is concluded from the experimental results that the external vessel cooling highly affected the ablation configuration and the thermal behaviors of the vessel and the penetration. An increase in coolant mass flow rate from 0.047 to 0.152 kg/s had effects on the thermal behavior of the lower head vessel and penetration in the SUS-EXT tests. The LILAC analytical results on temperature distribution and ablation depth in the lower head vessel and penetration were very similar to the experimental results

Progress of ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Ioki, K., E-mail: Kimihiro.Ioki@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Bayon, A. [F4E, c/ Josep Pla, No. 2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Choi, C.H.; Daly, E.; Dani, S.; Davis, J.; Giraud, B.; Gribov, Y.; Hamlyn-Harris, C.; Jun, C.; Levesy, B. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Kim, B.C. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Kuzmin, E. [NTC “Sintez”, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); Le Barbier, R.; Martinez, J.-M. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Pathak, H. [ITER-India, A-29, GIDC Electronic Estate, Sector 25, Gandhinagar 382025 (India); Preble, J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Sa, J.W. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Terasawa, A.; Utin, Yu. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); and others

2013-10-15

Highlights: ► This covers the overall status and progress of the ITER vacuum vessel activities. ► It includes design, R and D, manufacturing and approval process of the regulators. ► The baseline design was completed and now manufacturing designs are on-going. ► R and D includes ISI, dynamic test of keys and lip-seal welding/cutting technology. ► The VV suppliers produced full-scale mock-ups and started VV manufacturing. -- Abstract: Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R and D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure.

Progress of ITER vacuum vessel

International Nuclear Information System (INIS)

Ioki, K.; Bayon, A.; Choi, C.H.; Daly, E.; Dani, S.; Davis, J.; Giraud, B.; Gribov, Y.; Hamlyn-Harris, C.; Jun, C.; Levesy, B.; Kim, B.C.; Kuzmin, E.; Le Barbier, R.; Martinez, J.-M.; Pathak, H.; Preble, J.; Sa, J.W.; Terasawa, A.; Utin, Yu.

2013-01-01

Highlights: ► This covers the overall status and progress of the ITER vacuum vessel activities. ► It includes design, R and D, manufacturing and approval process of the regulators. ► The baseline design was completed and now manufacturing designs are on-going. ► R and D includes ISI, dynamic test of keys and lip-seal welding/cutting technology. ► The VV suppliers produced full-scale mock-ups and started VV manufacturing. -- Abstract: Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R and D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure

33 CFR Appendix B to Part 155 - Determining and Evaluating Required Response Resources for Vessel Response Plans

Science.gov (United States)

2010-07-01

... Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) POLLUTION OIL OR HAZARDOUS MATERIAL POLLUTION PREVENTION REGULATIONS FOR VESSELS Pt. 155, App. B Appendix B to Part 155—Determining and... environment. For example, vessels moving from the ocean to a river port must identify appropriate equipment...

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)

In-Vessel Coolability. Workshop Proceedings, in collaboration with EC-SARNET

International Nuclear Information System (INIS)

2011-01-01

Severe Accident Management Guidelines increase focus on containment integrity after some progression in the course of a severe accident. This change in priorities is made according to criteria that vary depending on reactor type and specific procedures. Once a water source has been recovered, different accident management strategies can be used: send water into the core and/or cool the reactor pressure vessel (RPV) externally. It should be noticed that, depending on the amount of water available, these strategies might conflict with other uses of water such as for instance activating spray systems in the containment or may have deleterious effects as for instance an increase in the production of hydrogen. Generally, for in-vessel reflooding, the models used for evaluation of accident management measures suffer from a lack of validation. Given this background, the objectives of the workshop were: -) to exchange information on different Severe Accident Management strategies used or contemplated for the in-vessel coolability issue; -) to review recent, ongoing and planned experimental programmes on reflooding; -) to review models used for reflooding in severe accident calculation tools, either simplified or sophisticated; -) to exchange information on the treatment of reflooding in different safety studies such as Probabilistic Safety Assessment; and -) to provide recommendations for future work, as necessary

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

Preliminary assessment for dust contamination of ITER In-Vessel Transporter

Energy Technology Data Exchange (ETDEWEB)

Saito, Makiko, E-mail: saito.makiko@jaea.go.jp [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan); Ueno, Kenichi; Maruyama, Takahito; Murakami, Shin; Takeda, Nobukazu; Kakudate, Satoshi [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan); Nakahira, Masataka; Tesini, Alessandro [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France)

2014-10-15

Highlights: •To assess the exposure to the maintenance workers, we calculated the effective dose rate. •To reduce the effective dose rate, the IVT was decontaminated and underwent a design change. •The effective dose rate at each maintenance point was also calculated. -- Abstract: After plasma operations of ITER, radioactive dust will have accumulated in the vacuum vessel (VV). The In-Vessel Transporter (IVT) will be introduced into the VV to remove the shield blanket modules for maintenance or replacement and later reinstall them. The IVT itself also needs to undergo regular maintenance in the Hot Cell Facility (HCF). It is assumed that maintenance workers will be exposed to radioactive dust that has adhered to the surfaces of the IVT. In this study, the areas of the IVT that may be contaminated by dust are evaluated to assess the level of exposure to workers during maintenance work in the HCF. Decontamination processes for the IVT, such as a combination of vacuuming and brushing, were investigated and the dose rate after these processes was evaluated. Even though dust was removed from surfaces where decontamination was possible, the dose rate was very high at some assessment points. To decrease the dose rate in accordance with ALARA policy, a decontamination plan and a maintenance plan, which includes the removal of dust, a radiation shield system, and a reduction in working time are proposed.

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.)

Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

46 CFR 154.1816 - Cargo location plan.

Science.gov (United States)

2010-10-01

... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo location plan. 154.1816 Section 154.1816 Shipping... FOR SELF-PROPELLED VESSELS CARRYING BULK LIQUEFIED GASES Operations § 154.1816 Cargo location plan. The master shall ensure that: (a) A cargo location plan is prepared that gives: (1) The location and...

Impact of chemistry on Standard High Solids Vessel Design mixing

Energy Technology Data Exchange (ETDEWEB)

Poirier, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-03-02

The plan for resolving technical issues regarding mixing performance within vessels of the Hanford Waste Treatment Plant Pretreatment Facility directs a chemical impact study to be performed. The vessels involved are those that will process higher (e.g., 5 wt % or more) concentrations of solids. The mixing equipment design for these vessels includes both pulse jet mixers (PJM) and air spargers. This study assesses the impact of feed chemistry on the effectiveness of PJM mixing in the Standard High Solids Vessel Design (SHSVD). The overall purpose of this study is to complement the Properties that Matter document in helping to establish an acceptable physical simulant for full-scale testing. The specific objectives for this study are (1) to identify the relevant properties and behavior of the in-process tank waste that control the performance of the system being tested, (2) to assess the solubility limits of key components that are likely to precipitate or crystallize due to PJM and sparger interaction with the waste feeds, (3) to evaluate the impact of waste chemistry on rheology and agglomeration, (4) to assess the impact of temperature on rheology and agglomeration, (5) to assess the impact of organic compounds on PJM mixing, and (6) to provide the technical basis for using a physical-rheological simulant rather than a physical-rheological-chemical simulant for full-scale vessel testing. Among the conclusions reached are the following: The primary impact of precipitation or crystallization of salts due to interactions between PJMs or spargers and waste feeds is to increase the insoluble solids concentration in the slurries, which will increase the slurry yield stress. Slurry yield stress is a function of pH, ionic strength, insoluble solids concentration, and particle size. Ionic strength and chemical composition can affect particle size. Changes in temperature can affect SHSVD mixing through its effect on properties such as viscosity, yield stress, solubility

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

Segmentation and packaging reactor vessels internals

International Nuclear Information System (INIS)

Boucau, Joseph

2014-01-01

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

RNL NDT studies related to PWR pressure vessel inlet nozzle inspection

International Nuclear Information System (INIS)

Rogerson, A.; Poulter, L.N.J.; Clough, P.; Cooper, A.

1984-01-01

Non-destructive examinations of the Reactor Pressure Vessel (RPV) of a Pressurized Water Reactor (PWR) play an important role in assuring vessel integrity throughout its operational life. Automated ultrasonic techniques for the detection and sizing of flaws in thick-section seam welds and near-surface regions in a PWR RPV have been under development at RNL for some time. Techniques for the inspection of complex geometry welds and other regions of the vessel are now being assessed and further developed as part of the UK NDT development programme in support of the Sizewell PWR. One objective of this programme is to demonstrate that the range of ultrasonic techniques already shown to be effective for the inspection of seam welds and inlet nozzle corner regions, through exercises such as the Defect Detection Trials, can also be effective for inspection of these other vessel regions. The nozzle-to-vessel welds and nozzle crotch corners associated with the RPV water inlet and outlet nozzles are two such regions being examined in this programme. In this paper, a review is given of the work performed at RNL in the development of a laboratory-based inspection system for inlet nozzle inspection. The main features of the system in its current stage of development are explained. (author)

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

Science.gov (United States)

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

2014-01-01

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

Study on the quality of FRP fishing vessel based on improved Fishbone Chart

Science.gov (United States)

Sui, J. H.; Yu, Y. F.; Du, Q. F.; Jiang, D. W.

2018-01-01

The construction quality of FRP fishing vessels influences their production, use and industry development. In order to explore the factors that affect the construction quality of FRP fishing vessels, key factors affecting the construction quality of FRP fishing vessels are determined based on the quality problems of FRP fishing vessels constructed. The improved Fishbone Chart is used to analyze the eight factors of “human, machine, material, process, environment, inspection, design and information”. Taking the factors that affect the construction quality of FRP fishing vessels as the central target, the eight influencing factors were condensed into five aspects and a composite Fishbone Chart is drawn. The Fishbone Chart is used as the basic model, the influencing factors are sorted, screened and discriminated, and the system model convenient for construction site management and control is established. Finally, the causes of poor construction of FRP fishing vessels are analyzed and discussed, and relevant suggestions are put forward.

Are the anatomy textbooks wrong? A clinical patho-anatomic study of foveal vessels in the round ligament of the hip.

Science.gov (United States)

Barrington, John W; Sodhi, Nipun; Ali, Muzaffar; Khlopas, Anton; Sultan, Assem A; Kee, James R; Holmes, Kristopher A; Hariri, Omar; Newman, Jared M; Mont, Michael A

2017-12-01

The foveal vessels of the ligamentum teres are an anterior branch of the posterior division of the obturator artery, providing blood to the capitis of the femoral head. However, the basic anatomic description of foveal vasculature in the ligamentum teres of the hip is widely variable, with some studies reporting that the vessels are not patent in roughly one third of all adults. Therefore, the purpose of this study was to evaluate the status of foveal vessels in primary total hip arthroplasty (THA) patients. Specifically, we evaluated: (I) if the foveal vessels were intact; and we (II) correlated foveal vessel status with (i) patient demographics, including gender and age; as well as (ii) perioperative data, such as operative time and blood loss. The macroscopic status of the foveal vessels in the ligamentum teres femoris was documented in 266 patients at the time of primary unilateral THA performed between August 2015 and April 2017. The vessels were considered to be intact if active bleeding was directly visible from the acetabular stump of the severed ligamentum teres femoris. Demographics including age, gender, and preoperative diagnosis were collected. The perioperative outcome variables included estimated blood loss (EBL) and operative time. Foveal vessel status defined as intact or not intact, was also correlated with patient demographics and perioperative data. A student's t -test was used to compare the continuous variables and a chi square test was used for categorical variables. The foveal vessels were intact in 161 patients (61%) and not intact in 105 patients (39%). The mean age for patients with intact foveal vessels was found to be 64 years (range, 18 to 94 years) vs. 65 years (range, 29 to 94 years) (P>0.05) for not intact. No correlation was found between preoperative diagnosis, gender, operative time, and EBL and foveal vessel status. The results of this clinical patho-anatomic study of the foveal vessels in the ligamentum teres femoris of the hip

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

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.

A study on ex-vessel steam explosion for a flooded reactor cavity of reactor scale - 15216

International Nuclear Information System (INIS)

Song, S.; Yoon, E.; Kim, Y.; Cho, Y.

2015-01-01

A steam explosion can occur when a molten corium is mixed with a coolant, more volatile liquid. In severe accidents, corium can come into contact with coolant either when it flows to the bottom of the reactor vessel and encounters the reactor coolant, or when it breaches the reactor vessel and flows into the reactor containment. A steam explosion could then threaten the containment structures, such as the reactor vessel or the concrete walls/penetrations of the containment building. This study is to understand the shortcomings of the existing analysis code (TEXAS-V) and to estimate the steam explosion loads on reactor scale and assess the effect of variables, then we compared results and physical phenomena. Sensitivity study of major parameters for initial condition is performed. Variables related to melt corium such as corium temperature, falling velocity and diameter of melt are more important to the ex-vessel steam explosion load and the steam explosion loads are proportional to these variables related to melt corium. Coolant temperature on reactor cavity has a specific area to increase the steam explosion loads. These results will be used to evaluate the steam explosion loads using ROAAM (Risk Oriented Accident Analysis Methodology) and to develop the evaluation methodology of ex-vessel steam explosion. (authors)

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)

Instrumentation of a prestressed concrete containment vessel model

International Nuclear Information System (INIS)

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

1995-01-01

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

Demarcation of inland vessels' limit off Mormugao port region, India: A pilot study for the safety of inland vessels using wave modelling

Digital Repository Service at National Institute of Oceanography (India)

Vethamony, P.; Aboobacker, V.M.; Sudheesh, K.; Babu, M.T.; AshokKumar, A.

The Ministry of Shipping desires to revise the inland vessels' limit (IVL) notification based on scientific rationale to improve the safety of vessels and onboard personnel. The Mormugao port region extending up to the Panaji was considered...

Deformable 4DCT lung registration with vessel bifurcations

International Nuclear Information System (INIS)

Hilsmann, A.; Vik, T.; Kaus, M.; Franks, K.; Bissonette, J.P.; Purdie, T.; Beziak, A.; Aach, T.

2007-01-01

In radiotherapy planning of lung cancer, breathing motion causes uncertainty in the determination of the target volume. Image registration makes it possible to get information about the deformation of the lung and the tumor movement in the respiratory cycle from a few images. A dedicated, automatic, landmark-based technique was developed that finds corresponding vessel bifurcations. Hereby, we developed criteria to characterize pronounced bifurcations for which correspondence finding was more stable and accurate. The bifurcations were extracted from automatically segmented vessel trees in maximum inhale and maximum exhale CT thorax data sets. To find corresponding bifurcations in both data sets we used the shape context approach of Belongie et al. Finally, a volumetric lung deformation was obtained using thin-plate spline interpolation and affine registration. The method is evaluated on 10 4D-CT data sets of patients with lung cancer. (orig.)

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.

Research vessels

Digital Repository Service at National Institute of Oceanography (India)

Rao, P.S.

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

A basic study on the ITER tritium storage vessel design and components

International Nuclear Information System (INIS)

Chung, H. S.; Ahn, D. H.; Kim, K. R.; Yim, S. P.; Paek, S. W.; Lee, M. S.; Lee, S. H.; Shim, M. H.

2006-01-01

The ZrCo getter beds are built of a primary vessel which contains the ZrCo powder mixed with Cu spheres of less than one mm diameter and of a secondary outer vessel. The purpose of the secondary outer vessel is to capture permeated or leaked tritium and to present a good thermal insulation when properly evacuated. A third volume, a helium filled loop, is installed in the primary volume to remove the decay heat and is used to perform tritium accountancy measurements

Inspection accessibility study of the Millstones Unit 1 Reactor Pressure Vessel

International Nuclear Information System (INIS)

Calhoun, G.; Kapoor, A.; Davis, J.B.

1990-07-01

The need for more extensive ultrasonic examination of the Reactor Pressure Vessel (RPV) shell welds at some of the BWR plants is becoming apparent. Recent NRC comments regarding limits on future inspection relief requests and proposed revisions to the ASME Boiler and Pressure Vessel Code, Section 11 have caused BWR plant owners and vendors of inspection services to look at inspection tooling that would allow for the ultrasonic examination of RPV shell welds previously considered inaccessible. One approach to inspection of the RPV shell welds that will allow for greater coverage at some BWR plants is to perform these inspections from inside the vessel. In late 1988, Westinghouse received a contract from Northeast Utilities to examine the flange ligament areas and the shell welds in the upper portion of the Millstone Unit 1 PRV during an outage scheduled to begin in April 1989. This examination was performed with an inspection tool placed inside of the vessel and the UDRPS data acquisition system. The inspection tool was based on tooling used routinely at the Swedish and Finnish BWRs. The object of RP C105-1 was to use the information gathered prior to and during the Millstone Unit 1 examination and to perform an inspection accessibility study and prepare a report covering the lessons learned during the examination that would benefit other BWR utilities considering similar inspections. The final report consists of two volumes. This document, Volume 1, describes a 3D model of the Millstone Unit 1 PRV that was based on information obtained from plant drawings. The model is a useful tool for visualizing areas of limited access from both inside and outside of the RPV. 5 refs., 38 figs., 4 tabs

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

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.

Histomorphometric comparative study of blood vessels and their pattern in follicular cyst, odontogenic keratocyst, and ameloblastoma.

Science.gov (United States)

Seifi, Safora; Feizi, Farideh; Khafri, Thoraya; Aram, Mehrdad

2013-03-01

The present study aimed at assessment and histomorphometric analysis of intratumoral and peritumoral (cystic) blood vessels in odontogenic lesions and their pattern on their clinical behavior by immunohistochemistry and morphometry. In a descriptive and analytical cross-sectional study, 45 paraffin blocks of ameloblastoma, odontogenic keratocyst, and follicular cyst were selected and stained immunohistochemically for CD34. In each slide, images of 3 microscopic fields with the highest microvessel density in intratumoral and peritumoral (cystic) areas were captured at 40× magnification with attached camera system. Inner vascular diameter (IVD) and outer vascular diameter (OVD), cross-sectional area (CSA), and the wall thickness (WT) of the vessels were measured with Motic Plus 2 software. The vascular pattern in odontogenic lesions was analyzed. Outer vascular diameter, IVD, and CSA of the vessels in peritumoral (cystic) areas were greater in ameloblastoma than keratocyst (P = 0.001) and follicular cyst (P keratocyst and follicular cyst. Morphometric specifications of blood vessels (IVD, OVD, CSA) and their pattern in peritumoral (cystic) areas may influence the aggressive clinical behavior of ameloblastoma in comparison with keratocyst and follicular cyst.

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

International Nuclear Information System (INIS)

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

2011-01-01

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

JT-60SA vacuum vessel manufacturing and assembly

Energy Technology Data Exchange (ETDEWEB)

Masaki, Kei, E-mail: masaki.kei@jaea.go.jp [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Shibama, Yusuke K.; Sakurai, Shinji; Shibanuma, Kiyoshi; Sakasai, Akira [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan)

2012-08-15

Highlights: Black-Right-Pointing-Pointer The design of the JT-60SA vacuum vessel body was completed with the demonstration of manufacturing procedure by the mock-up fabrication of the 20 Degree-Sign upper half of VV. Black-Right-Pointing-Pointer The actual VV manufacturing has started since November 2009. Black-Right-Pointing-Pointer The first product of the VV 40 Degree-Sign sector was completed in May 2011. Black-Right-Pointing-Pointer A basic VV assembly scenario and procedure were studied to complete the 360 Degree-Sign VV including positioning method and joint welding. - Abstract: The JT-60SA vacuum vessel (VV) has a D-shaped poloidal cross section and a toroidal configuration with 10 Degree-Sign segmented facets. A double wall structure is adopted to ensure high rigidity at operational load and high toroidal one-turn resistance. The material is 316L stainless steel with low cobalt content (<0.05%). The design temperatures of the VV at plasma operation and baking are 50 Degree-Sign C and 200 Degree-Sign C, respectively. In the double wall, boric-acid water is circulated at plasma operation to reduce the nuclear heating of the superconducting magnets. For baking, nitrogen gas is circulated in the double wall after draining of the boric-acid water. The manufacturing of the VV started in November 2009 after a fundamental welding R and D and a trial manufacturing of 20 Degree-Sign upper half mock-up. The manufacturing of the first VV 40 Degree-Sign sector was completed in May 2011. A basic concept and required jigs of the VV assembly were studied. This paper describes the design and manufacturing of the vacuum vessel. A plan of VV assembly in torus hall is also presented.

Plan's CCCD approach - Country study PLAN-Bangladesh

NARCIS (Netherlands)

Uffelen, van G.J.

2009-01-01

PLAN Netherlands National Organisation has in coordination with PLAN International Headquarters commissioned an independent formative evaluation study to get systematic insight in the preconditions for appropriate functionality of Child Centred Community Development (CCCD) and to strengthen common

Experimental Study of Interactions Between Sub-oxidized Corium and Reactor Vessel Steel

International Nuclear Information System (INIS)

Bechta, S.V.; Khabensky, V.B.; Granovsky, V.S.; Krushinov, E.V.; Vitol, S.A.; Gusarov, V.V.; Almiashev, V.I.; Lopukh, D.B.; Tromm, W.; Miassoedov, A.; Bottomley, D.; Fischer, M.; Piluso, P.; Altstadt, E.; Willschutz, H.G.; Fichoti, F.

2006-01-01

One of the critical factors in the analysis of in-vessel melt retention is the vessel strength. It is, in particular, sensitive to the thickness of intact vessel wall, which, in its turn, depends on the thermal conditions and physicochemical interactions with corium. Physicochemical interaction of prototypic UO 2 -ZrO 2 -Zr corium melt and VVER vessel steel was examined during the 2. Phase of the ISTC METCOR Project. Rasplav-3 test facility was used for conducting four tests, in which the Zr oxidation degree and interaction front temperature were varied; in one of the tests, stainless steel was added to the melt. Direct experimental measurements and post-test analyses were used for determining corrosion kinetics and maximum corrosion depth (i.e. the physicochemical impact of corium on the cooled vessel steel specimens), as well as the steel temperature conditions during the interaction, and finally the structure and composition of crystallized ingots, including the interaction zone. The minimum temperature on the interaction front boundary, which determined its final position and maximum corrosion depth was ∼ 1090 deg. C. An empirical correlation for calculation of corrosion kinetics has been derived. (authors)

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

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).;

Vessels Route Planning Problem with Uncertain Data

Directory of Open Access Journals (Sweden)

Tomasz Neumann

2016-09-01

Full Text Available The purpose of this paper is to find a solution for route planning in a transport networks, where the costs of tracks, factor of safety and travel time are ambiguous. This approach is based on the Dempster-Shafer theory and well known Dijkstra's algorithm. In this approach important are the influencing factors of the mentioned coefficients using uncertain possibilities presented by probability intervals. Based on these intervals the quality intervals of each route can be determined. Applied decision rules can be described by the end user.

Rapid sealing of porcine renal blood vessels, ex vivo, using a high power, 1470-nm laser, and laparoscopic prototype

Science.gov (United States)

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

2017-05-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

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

46 CFR 199.630 - Alternatives for passenger vessels in a specified service.

Science.gov (United States)

2010-10-01

... operating area including— (i) The scope and degree of the risks or hazards to which the vessel will be... commercial traffic; the presence of any unusual cargoes; and other similar factors; (iii) The port and...; and (iv) Environmental factors. (2) A comprehensive shipboard safety management and contingency plan...

Positron annihilation and Moessbauer studies of neutron irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

Brauer, G.; Matz, W.; Liszkay, L.; Molnar, B.

1990-11-01

Positron annihilation (lifetime, Doppler broadening) and Moessbauer studies on unirradiated, neutron irradiated and neutron irradiated plus annealed reactor pressure vessel steels (Soviet type 15Kh2NMFA) are presented. The role of microstructural properties and the formation of irradiation-induced precipitates is discussed. (orig.) [de

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

LSC (X) Logistics Planning Support, Task 1 Requirements Definition

National Research Council Canada - National Science Library

Scalzo, John

2004-01-01

.... During the early stage after vessel delivery, there is a requirement to manage this plan through a transition period until the designated support manager is under contract and/or provided with the resources to execute the plan.

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...

[Do double gloves protect against contamination during cannulation of blood vessels? A prospective randomized study].

Science.gov (United States)

Szarpak, Łukasz; Kurowski, Andrzej

2014-01-01

Undamaged medical gloves protect medical personnel from contact with physiological fluids of the patient. Thus they protect the assistance provider from hand skin contamination with potentially infectious biological materials. The aim of the study was to evaluate the occurrence of pierce, perforations or damage of medical gloves during cannulation of blood vessels. In the prospective randomized study 303 pairs of gloves, used during cannulation of blood vessels under simulated resuscitation, were analyzed. Gloves were tested by the water leak test. The water test revealed 44 cases of damage to the gloves used during cannulation of blood vessels. Significant differences were noted in the frequency of damage to both the outer and single pairs of gloves and the inner pair of gloves. The study showed that the use of double gloves provides a higher level of security for a paramedic than the use of a single pair of gloves, however, double gloves reduce the manual dexterity of a paramedic. A large number of damages to gloves are not noticed by medical personnel during surgery.

Probabilistic study of PWR reactor pressure vessel fracture

International Nuclear Information System (INIS)

Dufresne, J.; Lucia, A.C.; Grandemange, J.; Pellissier-Tanon, A.

1983-01-01

Different methods are used to evaluate the rupture probability of a nuclear pressure vessel. On of them extrapolates to nuclear pressure vessels, data of failure found in conventional pressure vessels. The disadvantage of such an approach is that the effects of systematic changes in key parameters cannot be taken into account. For example, the influence of irradiation and the use of quality assurance programs encompassing design, fabrication and materials cannot be considered. But the most important disadvantage of this method is the limited size of the representative population and consequently the high value of the upper bound failure rate corresponding to a requested confidence level. The method used in the present work involves the development of physical models based on an understanding of the failure modes and expressing the conventional concepts of fracture mechanics in a probabilistic form; the fatigue crack growth rate, calculated for conditions of cyclic loading, the initiation of unstable crack propagation, and the possibility of crack arrest. The analysis therefore requires the statistical expression of the factors and parameters which appear in the expressions of the law of crack growth and of toughness, and also those which are used in the calculation of the stress intensity factor K 1 . All input data are entered in COVASTOL code in histogram form. This code takes into account the degree of correlation between the flaw size and the Paris' law coefficients. It computes the propagation of a given defect in a given position, and the corresponding failure probability during accidental loading

40 CFR 300.210 - Federal contingency plans.

Science.gov (United States)

2010-07-01

... contingency plans under the national response system: The National Contingency Plan, RCPs, and ACPs. These... discharge under § 300.324, and to mitigate or prevent a substantial threat of such a discharge, from a vessel, offshore facility, or onshore facility operating in or near the area. (2) The areas of...

Design and Structural Analysis for the Vacuum Vessel of Superconducting Tokamak JT-60SC

International Nuclear Information System (INIS)

Kudo, Y.; Sakurai, S.; Masaki, K.; Urata, K.; Sasajima, T.; Matsukawa, M.; Sakasai, A.; Ishida, S.

2003-01-01

A modification of the JT-60 is planned to be a superconducting tokamak (JT-60SC) in order to establish steady-state operation of high beta plasma for 100 s, and to ensure the applicability of ferritic steel as a reduced activation material for reactor relevant break-even class plasmas. This paper describes the detailed design of the vacuum vessel, which has a unique structure for cost effective manufacturing, as well as structural analysis results for a feasibility study

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

Roi Detection and Vessel Segmentation in Retinal Image

Science.gov (United States)

Sabaz, F.; Atila, U.

2017-11-01

Diabetes disrupts work by affecting the structure of the eye and afterwards leads to loss of vision. Depending on the stage of disease that called diabetic retinopathy, there are sudden loss of vision and blurred vision problems. Automated detection of vessels in retinal images is a useful study to diagnose eye diseases, disease classification and other clinical trials. The shape and structure of the vessels give information about the severity of the disease and the stage of the disease. Automatic and fast detection of vessels allows for a quick diagnosis of the disease and the treatment process to start shortly. ROI detection and vessel extraction methods for retinal image are mentioned in this study. It is shown that the Frangi filter used in image processing can be successfully used in detection and extraction of vessels.

An experimental study of hypervapotron structure in external reactor vessel cooling

Energy Technology Data Exchange (ETDEWEB)

Zhao, Yufeng; Zhang, Ming [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei, E-mail: chenpeipei@snptc.com.cn [State Power Investment Group Corporation, Beijing (China)

2016-07-15

Highlights: • Experiments are performed to study the application of hypervapotron in ERVC design. • CHF experiments on two surfaces are conducted under different flow conditions. • Hypervapotron improves CHF performance by 40–60% compared with smooth surface. • Visualization shows fin structure removes vapor mushroom for better liquid supply. - Abstract: In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to

General and crevice corrosion study of the in-wall shielding materials for ITER vacuum vessel

Science.gov (United States)

Joshi, K. S.; Pathak, H. A.; Dayal, R. K.; Bafna, V. K.; Kimihiro, Ioki; Barabash, V.

2012-11-01

Vacuum vessel In-Wall Shield (IWS) will be inserted between the inner and outer shells of the ITER vacuum vessel. The behaviour of IWS in the vacuum vessel especially concerning the susceptibility to crevice of shielding block assemblies could cause rapid and extensive corrosion attacks. Even galvanic corrosion may be due to different metals in same electrolyte. IWS blocks are not accessible until life of the machine after closing of vacuum vessel. Hence, it is necessary to study the susceptibility of IWS materials to general corrosion and crevice corrosion under operations of ITER vacuum vessel. Corrosion properties of IWS materials were studied by using (i) Immersion technique and (ii) Electro-chemical Polarization techniques. All the sample materials were subjected to a series of examinations before and after immersion test, like Loss/Gain weight measurement, SEM analysis, and Optical stereo microscopy, measurement of surface profile and hardness of materials. After immersion test, SS 304B4 and SS 304B7 showed slight weight gain which indicate oxide layer formation on the surface of coupons. The SS 430 material showed negligible weight loss which indicates mild general corrosion effect. On visual observation with SEM and Metallography, all material showed pitting corrosion attack. All sample materials were subjected to series of measurements like Open Circuit potential, Cyclic polarization, Pitting potential, protection potential, Critical anodic current and SEM examination. All materials show pitting loop in OC2 operating condition. However, its absence in OC1 operating condition clearly indicates the activity of chloride ion to penetrate oxide layer on the sample surface, at higher temperature. The critical pitting temperature of all samples remains between 100° and 200°C.

33 CFR 155.1040 - Response plan requirements for unmanned tank barges carrying oil as a primary cargo.

Science.gov (United States)

2010-07-01

... POLLUTION PREVENTION REGULATIONS FOR VESSELS Tank Vessel Response Plans for Oil § 155.1040 Response plan...-based support or advice; (ii) The individuals who shall be notified of a casualty potentially affecting... coordinator responsible for overseeing or directing those actions. (4) The organizational structure that will...

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

International Nuclear Information System (INIS)

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

1981-08-01

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

33 CFR 155.1230 - Response plan development and evaluation criteria.

Science.gov (United States)

2010-07-01

... impact; (2) Oil recovery devices appropriate for the type of animal fats or vegetable oils carried; and (3) Other appropriate equipment necessary to respond to a discharge involving the type of animal fats... VESSELS Response plan requirements for vessels carrying animal fats and vegetable oils as a primary cargo...

Bone marrow blood vessels: normal and neoplastic niche

Directory of Open Access Journals (Sweden)

Saeid Shahrabi

2016-11-01

Full Text Available Blood vessels are among the most important factors in the transport of materials such as nutrients and oxygen. This study will review the role of blood vessels in normal bone marrow hematopoiesis as well as pathological conditions like leukemia and metastasis. Relevant literature was identified by a Pubmed search (1992-2016 of English-language papers using the terms bone marrow, leukemia, metastasis, and vessel. Given that blood vessels are conduits for the transfer of nutrients, they create a favorable situation for cancer cells and cause their growth and development. On the other hand, blood vessels protect leukemia cells against chemotherapy drugs. Finally, it may be concluded that the vessels are an important factor in the development of malignant diseases.

Power reactor embrittlement data base (PR-EDB): Uses in evaluating radiation embrittlement of reactor vessels

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

1992-01-01

Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current Codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed, computerized data base. Also, such a data is essential for the evaluation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current compilation contains data from 92 reactors and consists of 175 data points for weld materials (79 different welds) and 395 data points for base materials (110 different base materials). The different types of data that are implemented or planned for this data base are discussed. ''User-friendly'' utility programs have been written to investigate a list of problems using this data base. The utility programs are also used to add and upgrade data, retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in this paper

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)

Superior long term outcome associated with native vessel versus graft vessel PCI following secondary PCI in patients with prior CABG.

Science.gov (United States)

Mavroudis, Chrysostomos A; Kotecha, Tushar; Chehab, Omar; Hudson, Jonathan; Rakhit, Roby D

2017-02-01

Secondary percutaneous coronary intervention (PCI) in patients with prior coronary artery bypass graft surgery is increasingly common. Graft vessel PCI has higher rates of adverse events compared with native coronary vessel PCI. To investigate the clinical outcomes of patients with prior CABG who underwent secondary PCI of either a graft vessel (GV), a native coronary vessel (NV) or both graft and native (NG) vessels. 220 patients (84% male) who underwent PCI in our institution to either GV (n=89), NV (n=103) or both GV and NV (NG group) (n=28) were studied. The study population underwent 378 procedures (GV group; n=126, NV group; n=164 and NG group; n=88). Median follow up was for 36months [range 2-75months]. Target vessel revascularisation (TVR) occurred in 12.5% of the GV group and 3.6% in the NV group [p=0.0004], and was predominantly due to in-stent restenosis. Patients who had PCI due to TVR were more likely to suffer from diabetes and peripheral vascular disease. History of chronic renal failure was associated with higher risk (HR 2.21, p=0.005) whereas preserved left ventricular ejection fraction (LVEF) with lower risk (HR 0.17, p=0.0007) of death. The median survival (interval between CABG and end of follow-up period) was lower in the GV compared with the NV group (315 vs 372months p=0.005). This registry demonstrates inferior long term outcome for patients undergoing secondary PCI of GV versus NV. Where possible, a strategy of NV rather than GV target PCI should be considered in patients with prior CABG. Secondary PCI in patients with prior CABG surgery is increasingly common. Graft vessel PCI has inferior outcomes with high rates of restenosis and occlusion compared with native coronary vessel PCI. We studied the clinical outcomes of 220 patients with prior CABG who underwent secondary PCI to either a graft vessel (GV), a native coronary vessel (NV) or both graft and native (NG) vessels. Target vessel revascularisation was 5 times higher in the GV

Iodine-infused aeration for hull fouling prevention: a vessel-scale study.

Science.gov (United States)

Dickenson, Natasha C; Krumholz, Jason S; Hunsucker, Kelli Z; Radicone, Michael

2017-11-01

Biofouling is a significant economic and ecological problem, causing reduced vessel performance and increases in fuel consumption and emissions. Previous research has shown iodine vapor (I 2 )-infused aeration to be an environmentally friendly method for deterring the settlement of fouling organisms. An aeration system was deployed on a vessel with hull sections coated with two types of antifoulant coatings, Intersleek ® 1100 (fouling-release) and Interspeed ® BRA-640 (ablative copper biocide), as well as an inert epoxy barrier coating, to assess the effectiveness of aeration in conjunction with common marine coatings. I 2 -infused aeration resulted in consistent reductions of 80-90% in hard fouling across all three coatings. Additionally, aeration reduced the soft fouling rate by 45-70% when used in conjunction with both Intersleek ® and Interspeed ® BRA versus those coatings alone. The results of this study highlight the contribution of I 2 -infused aeration as a standalone mechanism for fouling prevention or as a complement to traditional antifouling coatings.

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

ROI DETECTION AND VESSEL SEGMENTATION IN RETINAL IMAGE

Directory of Open Access Journals (Sweden)

F. Sabaz

2017-11-01

Full Text Available Diabetes disrupts work by affecting the structure of the eye and afterwards leads to loss of vision. Depending on the stage of disease that called diabetic retinopathy, there are sudden loss of vision and blurred vision problems. Automated detection of vessels in retinal images is a useful study to diagnose eye diseases, disease classification and other clinical trials. The shape and structure of the vessels give information about the severity of the disease and the stage of the disease. Automatic and fast detection of vessels allows for a quick diagnosis of the disease and the treatment process to start shortly. ROI detection and vessel extraction methods for retinal image are mentioned in this study. It is shown that the Frangi filter used in image processing can be successfully used in detection and extraction of vessels.

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...

Fast Generation of Container Vessel Stowage Plans:Using mixed integer programming for optimal master planning and constraint based local search for slot planning

OpenAIRE

Pacino, Dario

2012-01-01

Containerization has changed the way the world perceives shipping. It is now possible to establish complex international supply chains that have minimized shipping costs. Over the past two decades, the demand for cost efficient containerized transportation has seen a continuous increase. In order to answer to this demand, shipping companies have deployed bigger container vessels, that nowadays can transport up to 18,000 containers and are wider than the extended Panama Canal. Like busses, con...

Evaluation of In-Vessel Corium Retention under a Severe Accident

Energy Technology Data Exchange (ETDEWEB)

Park, Rae-Joon; Kang, Kyung-Ho; Ha, Kwang-Soon; Kim, Jong-Tae; Koo, Kil-Mo; Cho, Young-Ro; Hong, Seong-Wan; Kim, Sang-Baik; Kim, Hee-Dong

2008-02-15

The current study on In-Vessel corium Retention and its application activities to the actual nuclear power plant have been reviewed and discussed in this study. Severe accident sequence which determines an initial condition of the IVR has been evaluated and late phase melt progression, heat transfer on the outer reactor vessel, and in-vessel corium cooling mechanism have been estimated in detail. During the high pressure sequence of the reactor coolant system, a natural circulation flow of the hot steam leads to a failure of the pressurizer surge line before the reactor vessel failure, which leads to a rapid decrease of the reactor coolant system pressure. The results of RASPLAV/MASCA study by OECD/NEA have shown that a melt stratification has occurred in the lower plenum of the reactor vessel. In particular, laver inversion has occurred, which is that a high density of the metal melt moves to the lower part of the oxidic melt layer. A method of heat transfer enhancement on the outer reactor vessel is an optimal design of the reactor vessel insulation for an increase of the natural circulation flow between the outer reactor vessel and the its insulation, and an increase of the critical Heat flux on the outer reactor vessel by using various method, such as Nono fluid, coated reactor vessel, and so on. An increase method of the in-vessel melt cooling is a development of the In-vessel core catcher and a decrease of focusing effect in the metal layer.

QFD-ANP Approach for the Conceptual Design of Research Vessels: A Case Study

Science.gov (United States)

Venkata Subbaiah, Kambagowni; Yeshwanth Sai, Koneru; Suresh, Challa

2016-10-01

Conceptual design is a subset of concept art wherein a new idea of product is created instead of a visual representation which would directly be used in a final product. The purpose is to understand the needs of conceptual design which are being used in engineering designs and to clarify the current conceptual design practice. Quality function deployment (QFD) is a customer oriented design approach for developing new or improved products and services to enhance customer satisfaction. House of quality (HOQ) has been traditionally used as planning tool of QFD which translates customer requirements (CRs) into design requirements (DRs). Factor analysis is carried out in order to reduce the CR portions of HOQ. The analytical hierarchical process is employed to obtain the priority ratings of CR's which are used in constructing HOQ. This paper mainly discusses about the conceptual design of an oceanographic research vessel using analytical network process (ANP) technique. Finally the QFD-ANP integrated methodology helps to establish the importance ratings of DRs.

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.

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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

Platform image processing to study the structural properties of retinal vessel

Directory of Open Access Journals (Sweden)

Miguel Ángel MERCHÁN

2013-05-01

Full Text Available This paper presents a technological platform specialized in assessing retinal vessel caliber and describing the relationship of the results obtained to cardiovascular risk. Retinal circulation is an area of active research by numerous groups, and there is general experimental agreement on the analysis of the patterns of the retinal blood vessels in the normal human retina. The development of automated tools designed to improve performance and decrease interobserver variability, therefore, appears necessary. 

Gamma dose rate estimation and operation management suggestions for decommissioning the reactor pressure vessel of HTR-PM

Energy Technology Data Exchange (ETDEWEB)

Sheng Fang; Hong Li; Jianzhu Cao; Wenqian Li; Feng Xie; Jiejuan Tong [Institute of Nuclear and New Energy Technology, Tsinghua, University, Beijing (China)

2013-07-01

China is now designing and constructing a high temperature gas cooled reactor-pebble bed module (HTR-PM). In order to investigate the future decommissioning approach and evaluate possible radiation dose, gamma dose rate near the reactor pressure vessel was calculated for different cooling durations using QAD-CGA program. The source term of this calculation was provided by KORIGEN program. Based on the calculated results, the spatial distribution and temporal changes of gamma dose rate near reactor pressure vessel was systematically analyzed. A suggestion on planning decommissioning operation of reactor pressure vessel of HTRPM was given based on calculated dose rate and the Chinese Standard GB18871-2002. (authors)

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.

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

Development of a Remote Handling Robot for the Maintenance of an ITER-Like D-Shaped Vessel

Directory of Open Access Journals (Sweden)

Peihua Chen

2014-01-01

Full Text Available Robotic operation is one of the major challenges in the remote maintenance of ITER vacuum vessel (VV and future fusion reactors as inner operations of Tokamak have to be done by robots due to the internal adverse conditions. This paper introduces a novel remote handling robot (RHR for the maintenance of ITER-like D-shaped vessel. The modular designed RHR, which is an important part of the remote handling system for ITER, consists of three parts: an omnidirectional transfer vehicle (OTV, a planar articulated arm (PAA, and an articulated teleoperated manipulator (ATM. The task of RHR is to carry processing tools, such as the viewing system, leakage detector, and electric screwdriver, to inspect and maintain the components installed inside the D-shaped vessel. The kinematics of the OTV, as well as the kinematic analyses of the PAA and ATM, is studied in this paper. Because of its special length and heavy payload, the dynamics of the PAA is also investigated through a dynamic simulation system based on robot technology middleware (RTM. The results of the path planning, workspace simulations, and dynamic simulation indicate that the RHR has good mobility together with satisfying kinematic and dynamic performances and can well accomplish its maintenance tasks in the ITER-like D-shaped vessel.

Power reactor pressure vessel benchmarks

International Nuclear Information System (INIS)

Rahn, F.J.

1978-01-01

A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

Advanced Approach of Reactor Pressure Vessel In-service Inspection

International Nuclear Information System (INIS)

Matokovic, A.; Picek, E.; Pajnic, M.

2006-01-01

The most important task of every utility operating a nuclear power plant is the continuously keeping of the desired safety and reliability level. This is achieved by the performance of numerous inspections of the components, equipment and system of the nuclear power plant in operation and in particular during the scheduled maintenance periods at re-fueling time. Periodic non-destructive in-service inspections provide most relevant criteria of the integrity of primary circuit pressure components. The task is to reliably detect defects and realistically size and characterize them. One of most important and the most extensive examination is a reactor pressure vessel in-service inspection. That inspection demand high standards of technology and quality and continual innovation in the field of non-destructive testing (NDT) advanced technology as well as regarding reactor pressure vessel tool and control systems. A remote underwater contact ultrasonic technique is employed for the examination of the defined sections (reactor welds), whence eddy current method is applied for clad surface examinations. Visual inspection is used for examination of the vessel inner surface. The movement of probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with NDT systems. The successful performance is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen NDT techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state of the art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. The advanced approach as presented in this paper offer more flexibility of application (non-destructive tests, local grinding action as well as taking of boat samples

Reactors with pressure vessel in pre-stressed concrete

International Nuclear Information System (INIS)

Devillers, Christian; Lafore, Pierre

1964-12-01

After having proposed a general description of the evolution of the general design of reactors with a vessel in pre-stressed concrete, this report outlines the interest of this technical solution of a vessel in pre-stressed concrete with integrated exchangers, which is to replace steel vessel. This solution is presented as much safer. The authors discuss the various issues related to protection: inner and outer biological protection of the vessel, material protection (against heating, steel irradiation, Wigner effect, and moderator radiolytic corrosion). They report the application of calculation methods: calculation of vessel concrete heating, study of the intermediate zone in integrated reactors, neutron spectrum and flows in the core of a graphite pile

Relationship between retinal vessel diameters and retinopathy in the Inter99 Eye Study

DEFF Research Database (Denmark)

Drobnjak, Dragana; Munch, Inger Christine; Glümer, Charlotte

2017-01-01

PURPOSE: To examine the association between retinal vessel diameters and retinopathy in participants with and without type 2 diabetes in a Danish population-based cohort. METHODS: The study included 878 persons aged 30 to 60 years from the Inter99 Eye Study. Retinopathy was defined as a presence...... was 6.3 µm (CI 95%: 1.0 to 11.6, p = 0.020) wider and CRVE was 7.9 µm (CI 95%: 0.7 to 15.2, p = 0.030) wider in those with retinopathy compared to those without retinopathy, after adjusting for age, gender, HbA1c, blood pressure, smoking, serum total and HDL cholesterol. In all participants, CRAE.......001), and decreased with higher HDL cholesterol (p gender was associated with wider CRVE (p = 0.029). CONCLUSIONS: Wider retinal vessel diameters were associated with the presence of retinopathy in participants with diabetes, but not in participants without diabetes...

Studies on structural analysis related to the design of the JT-60 vacuum vessel

International Nuclear Information System (INIS)

Takatsu, Hideyuki

1987-06-01

Studies on structural analysis of a vacuum vessel of tokamak-type fusion devices are presented. The present studies are proposals for the structural analysis procedures of the tokamak-type fusion devices and are composed of five parts, each of which covers the fundamental area required for the structural analysis and design; stress analysis, dynamic response analysis, fatigue evaluation, buckling analysis and seismic analysis. Special attention is paid to the critical component, bellows and the critical load, electromagnetic forces. A new finite element method modeling technique is proposed for the stress analysis of U-shaped bellows, where the bellows is replaced by an orthotropic plate having the same stiffness as the bellows. The applicability of the present modeling technique is confirmed by verification tests. Dynamic response and fatigue of the vacuum vessel are critical issues of the structural analysis and design of the tokamak-type fusion devices. Detailed dynamic response analyses of the JT-60 vacuum vessel are presented paying special attention to the dynamic behavior of the U-shaped bellows, where the above-mentioned modeling technique of the U-shaped bellows is applied. A fatigue evaluation method of the vacuum vessel under the dynamic electromagnetic forces is proposed, which utilizes the results of the detailed dynamic response analysis. In the present method, fatigue evaluation method for random loads is applied. Torsional fatigue strength of the welded bellows is experimentally evaluated aiming the application to the port of the fusion device and it is shown that the welded bellows reveals elastic buckling and spiral distortion under a small angle of tortion. Two formulae are proposed to evaluate the stress of the welded bellows under the forced angle of tortion. (author)

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.)

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

How to replace a reactor pressure vessel

International Nuclear Information System (INIS)

Huber, R.

1996-01-01

A potential life extending procedure for a nuclear reactor after, say, 40 years of service life, might in some circumstances be the replacement of the reactor pressure vessel. Neutron induced degradation of the vessel might make replacement by one of a different material composition desirable, for example. Although the replacement of heavy components, such as steam generators, has been possible for many years, the pressure vessel presents a much more demanding task if only because it is highly irradiated. Some preliminary feasibility studies by Siemens are reported for the two removal strategies that might be considered. These are removal of the entire pressure vessel in one piece and dismantling it into sections. (UK)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

Effective operational oil spill response planning

International Nuclear Information System (INIS)

Meyers, R.J.

1991-01-01

An operational Contingency Plan is one of the single most important aspects of effective oil spill response operations. It is a spill control game plan. A thorough contingency plan provides a set of guidelines that can be used to help direct all phases of spill response activities. More than simple a compilation of lists and rosters, the contingency plan reflects strategic and philosophical elements of spill response that help to ensure a viable response to any spill incident. Facilities and oil carrying vessels should have well maintained contingency plans with these features. This paper describes the requirement for effective oil spill response pans and the training required to exercise them

A Measure of Similarity Between Trajectories of Vessels

Directory of Open Access Journals (Sweden)

Le QI

2016-03-01

Full Text Available The measurement of similarity between trajectories of vessels is one of the kernel problems that must be addressed to promote the development of maritime intelligent traffic system (ITS. In this study, a new model of trajectory similarity measurement was established to improve the data processing efficiency in dynamic application and to reflect actual sailing behaviors of vessels. In this model, a feature point detection algorithm was proposed to extract feature points, reduce data storage space and save computational resources. A new synthesized distance algorithm was also created to measure the similarity between trajectories by using the extracted feature points. An experiment was conducted to measure the similarity between the real trajectories of vessels. The growth of these trajectories required measurements to be conducted under different voyages. The results show that the similarity measurement between the vessel trajectories is efficient and correct. Comparison of the synthesized distance with the sailing behaviors of vessels proves that results are consistent with actual situations. The experiment results demonstrate the promising application of the proposed model in studying vessel traffic and in supplying reliable data for the development of maritime ITS.

Planning for study abroad.

Science.gov (United States)

Freeman, L H; Hermann, C P

1994-01-01

There is more to planning for study abroad than buying airplane tickets. Providing a successful course away from the host institution requires a great deal of planning by the educator. This planning includes course development, budget, marketing, travel, housing, food, and entertainment. Careful preparation provides both the educator and student a unique and enriching experience that goes beyond the traditional course.

Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

International Nuclear Information System (INIS)

Mizutani, J.; Kawamura, S.; Aoki, M.; Mori, T.

2001-01-01

Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

Investigation of vessel visibility of iterative reconstruction method in coronary computed tomography angiography using simulated vessel phantom

International Nuclear Information System (INIS)

Inoue, Takeshi; Uto, Fumiaki; Ichikawa, Katsuhiro; Hara, Takanori; Urikura, Atsushi; Hoshino, Takashi; Miura, Youhei; Terakawa, Syouichi

2012-01-01

Iterative reconstruction methods can reduce the noise of computed tomography (CT) images, which are expected to contribute to the reduction of patient dose CT examinations. The purpose of this study was to investigate impact of an iterative reconstruction method (iDose 4 , Philips Healthcare) on vessel visibility in coronary CT angiography (CTA) by using phantom studies. A simulated phantom was scanned by a CT system (iCT, Philips Healthcare), and the axial images were reconstructed by filtered back projection (FBP) and given a level of 1 to 7 (L1-L7) of the iterative reconstruction (IR). The vessel visibility was evaluated by a quantitative analysis using profiles across a 1.5-mm diameter simulated vessel as well as visual evaluation for multi planar reformation (MPR) images and volume rendering (VR) images in terms of the normalized-rank method with analysis of variance. The peak CT value of the profiles decreased with IR level and full width at half maximum of the profile also decreased with the IR level. For normalized-rank method, there was no statistical difference between FBP and L1 (20% dose reduction) for both MPR and VR images. The IR levels higher than L1 sacrificed the spatial resolution for the 1.5-mm simulated vessel, and their visual vessel visibilities were significantly inferior to that of the FBP. (author)

Molecular magnetic resonance imaging of atherosclerotic vessel wall disease

Energy Technology Data Exchange (ETDEWEB)

Noerenberg, Dominik [Charite - University Medicine Berlin, Department of Radiology, Berlin (Germany); University of Munich - Grosshadern, Department of Clinical Radiology, Munich (Germany); Ebersberger, Hans U. [Heart Center Munich-Bogenhausen, Department of Cardiology and Intensive Care Medicine, Munich (Germany); Diederichs, Gerd; Hamm, Bernd [Charite - University Medicine Berlin, Department of Radiology, Berlin (Germany); Botnar, Rene M. [King' s College London, Division of Imaging Sciences and Biomedical Engineering, London (United Kingdom); Makowski, Marcus R. [Charite - University Medicine Berlin, Department of Radiology, Berlin (Germany); King' s College London, Division of Imaging Sciences and Biomedical Engineering, London (United Kingdom)

2016-03-15

Molecular imaging aims to improve the identification and characterization of pathological processes in vivo by visualizing the underlying biological mechanisms. Molecular imaging techniques are increasingly used to assess vascular inflammation, remodeling, cell migration, angioneogenesis and apoptosis. In cardiovascular diseases, molecular magnetic resonance imaging (MRI) offers new insights into the in vivo biology of pathological vessel wall processes of the coronary and carotid arteries and the aorta. This includes detection of early vascular changes preceding plaque development, visualization of unstable plaques and assessment of response to therapy. The current review focuses on recent developments in the field of molecular MRI to characterise different stages of atherosclerotic vessel wall disease. A variety of molecular MR-probes have been developed to improve the non-invasive detection and characterization of atherosclerotic plaques. Specifically targeted molecular probes allow for the visualization of key biological steps in the cascade leading to the development of arterial vessel wall lesions. Early detection of processes which lead to the development of atherosclerosis and the identification of vulnerable atherosclerotic plaques may enable the early assessment of response to therapy, improve therapy planning, foster the prevention of cardiovascular events and may open the door for the development of patient-specific treatment strategies. (orig.)

Molecular magnetic resonance imaging of atherosclerotic vessel wall disease

International Nuclear Information System (INIS)

Noerenberg, Dominik; Ebersberger, Hans U.; Diederichs, Gerd; Hamm, Bernd; Botnar, Rene M.; Makowski, Marcus R.

2016-01-01

Molecular imaging aims to improve the identification and characterization of pathological processes in vivo by visualizing the underlying biological mechanisms. Molecular imaging techniques are increasingly used to assess vascular inflammation, remodeling, cell migration, angioneogenesis and apoptosis. In cardiovascular diseases, molecular magnetic resonance imaging (MRI) offers new insights into the in vivo biology of pathological vessel wall processes of the coronary and carotid arteries and the aorta. This includes detection of early vascular changes preceding plaque development, visualization of unstable plaques and assessment of response to therapy. The current review focuses on recent developments in the field of molecular MRI to characterise different stages of atherosclerotic vessel wall disease. A variety of molecular MR-probes have been developed to improve the non-invasive detection and characterization of atherosclerotic plaques. Specifically targeted molecular probes allow for the visualization of key biological steps in the cascade leading to the development of arterial vessel wall lesions. Early detection of processes which lead to the development of atherosclerosis and the identification of vulnerable atherosclerotic plaques may enable the early assessment of response to therapy, improve therapy planning, foster the prevention of cardiovascular events and may open the door for the development of patient-specific treatment strategies. (orig.)

Predicting Human Mobility Patterns in Marine Ecosystems: Entropy and Home Range Calculations Based on High-Resolution Fishing Vessel Tracking Data

Science.gov (United States)

Murawski, S. A.

2016-02-01

A number of recent studies have developed metrics of human mobility patterns based on georeferenced cell phone records. The studies generally indicate a high degree of predictability in human location and relatively narrow home ranges for most people. In marine ecosystems there are a number of important uses for such calculations including marine spatial planning and predicting the impacts of marine management options such as establishing marine protected areas (MPAs). In this study we use individual fishing vessel satellite tracking (VMS) records ( 30 million records) obtained from commercial reef fish fishing vessels in the Gulf of Mexico during 2006-2014. This period witnessed the establishment of a variety of new regulations including individual fishing quotas (IFQs) for snapper, grouper, and tilefish, establishment of spatial-area closures, and the temporary closure of as much as 85,000 nautical miles of productive fishing grounds associated with the Deepwater Horizon oil spill accident. Vessel positions were obtained, with a location frequency of one hour. From these VMS data we calculated three measures of entropy (degree of repeatability in spatial use), as well as calculated the axis of gyration (home range) for each vessel in the data set. These calculations were related to a variety of descriptor variables including vessel size, distance from home port to predominant fishing grounds, revenue generated on fishing trips, and fishing regulations. The applicability of these calculations to marine resource management applications is discussed.

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

An integrated CAD/CAM system for CNG pressure vessel manufactured by deep drawing and ironing operation

Energy Technology Data Exchange (ETDEWEB)

Park, Joon Hong; Kim, Chul; Choi, Jae Chan [Pusan National Univ., Pusan (Korea, Republic of)

2004-06-01

The fiber reinforced composite material is widely used in the multi-industrial field because of their high specific modulus and specific strength. It has two main merits which are to cut down energy by reducing weight and to prevent explosive damage proceeding to the sudden bursting which is generated by the pressure leakage condition. Therefore, pressure vessels using this composite material can be applied in the field such as defence industry and aerospace industry. In this paper, for nonlinear finite element analysis of E-glass/epoxy filament winding of composite vessel subjected to internal pressure, the standard interpretation model is developed by using the ANSYS with AutoLISP and ANSYS APDL languages, general commercial software, which is verified as useful characteristic of the solution. Among the modules of the system, both the process planning module for carrying out the process planning of filament wound composite pressure vessel and the autofrettage process module for obtaining higher residual stress will minimize trial and error and reduce the period for developing new products. The system can serve as a valuable system for experts and as a dependable training aid for beginners.

An integrated CAD/CAM system for CNG pressure vessel manufactured by deep drawing and ironing operation

International Nuclear Information System (INIS)

Park, Joon Hong; Kim, Chul; Choi, Jae Chan

2004-01-01

The fiber reinforced composite material is widely used in the multi-industrial field because of their high specific modulus and specific strength. It has two main merits which are to cut down energy by reducing weight and to prevent explosive damage proceeding to the sudden bursting which is generated by the pressure leakage condition. Therefore, pressure vessels using this composite material can be applied in the field such as defence industry and aerospace industry. In this paper, for nonlinear finite element analysis of E-glass/epoxy filament winding of composite vessel subjected to internal pressure, the standard interpretation model is developed by using the ANSYS with AutoLISP and ANSYS APDL languages, general commercial software, which is verified as useful characteristic of the solution. Among the modules of the system, both the process planning module for carrying out the process planning of filament wound composite pressure vessel and the autofrettage process module for obtaining higher residual stress will minimize trial and error and reduce the period for developing new products. The system can serve as a valuable system for experts and as a dependable training aid for beginners

Thermodynamic study on the in-vessel corium - Application to the corium/concrete interaction

International Nuclear Information System (INIS)

Quaini, Andrea

2015-01-01

During a severe accident in a pressurised water reactor, the nuclear fuel can interact with the Zircaloy cladding, the neutronic absorber and the surrounding metallic structure forming a partially or completely molten mixture. The molten core can then interact with the reactor steel vessel forming a mixture called in-vessel corium. In the worst case, this mixture can pierce the vessel and pour onto the concrete underneath the reactor, leading the formation of the ex-vessel corium. Furthermore, depending on the considered scenario, the corium can be formed by a liquid phase or by two liquids, one metallic the other oxide. The objective of this thesis is the investigation of the thermodynamics of the prototypic in-vessel corium U-Pu-Zr- Fe-O. The approach used during the thesis is based on the CALPHAD method, which allows to obtain a thermodynamic model for this complex system starting from phase diagram and thermodynamic data. Heat treatments performed on the O-U-Zr system allowed to measure two tie-lines in the miscibility gap in the liquid phase at 2567 K. Furthermore, the liquidus temperatures of three Zr-enriched samples have been obtained by laser heating in collaboration with ITU. With the same laser heating technique, solidus temperatures have been obtained on the UO 2 -PuO 2 -ZrO 2 system. The influence of the reducing or oxidising on the melting behaviour of this system has been studied for the first time. The results show that the oxygen stoichiometry of these oxides strongly depends on the oxygen potential and on the metal composition of the samples. The miscibility gap in the liquid phase of the U-Zr-Fe-O system has been also observed. The whole set of experimental results with the literature data allowed to develop the thermodynamic model of the U-Pu-Zr-Fe-O system. Solidification path calculations have been performed for all the investigated samples to interpret the microstructures of the solidified samples. A good accordance has been obtained between

Quantification of Tumor Vessels in Glioblastoma Patients Using Time-of-Flight Angiography at 7 Tesla: A Feasibility Study

Science.gov (United States)

Radbruch, Alexander; Eidel, Oliver; Wiestler, Benedikt; Paech, Daniel; Burth, Sina; Kickingereder, Philipp; Nowosielski, Martha; Bäumer, Philipp; Wick, Wolfgang; Schlemmer, Heinz-Peter; Bendszus, Martin; Ladd, Mark; Nagel, Armin Michael; Heiland, Sabine

2014-01-01

Purpose To analyze if tumor vessels can be visualized, segmented and quantified in glioblastoma patients with time of flight (ToF) angiography at 7 Tesla and multiscale vessel enhancement filtering. Materials and Methods Twelve patients with newly diagnosed glioblastoma were examined with ToF angiography (TR = 15 ms, TE = 4.8 ms, flip angle = 15°, FOV = 160×210 mm2, voxel size: 0.31×0.31×0.40 mm3) on a whole-body 7 T MR system. A volume of interest (VOI) was placed within the border of the contrast enhancing part on T1-weighted images of the glioblastoma and a reference VOI was placed in the non-affected contralateral white matter. Automated segmentation and quantification of vessels within the two VOIs was achieved using multiscale vessel enhancement filtering in ImageJ. Results Tumor vessels were clearly visible in all patients. When comparing tumor and the reference VOI, total vessel surface (45.3±13.9 mm2 vs. 29.0±21.0 mm2 (pTesla MRI enables characterization and quantification of the internal vascular morphology of glioblastoma and may be used for the evaluation of therapy response within future studies. PMID:25415327

Experimental and theoretical studies on the high pressure vessel

International Nuclear Information System (INIS)

So, Dong Sup

1992-02-01

A High Pressure Melt Ejection (HPME) is one of the most important phenomena relevant to Direct Containment Heating(DCH) which could lead to an early containment failure in a several accident of PWRs. Dispersal of core debris following a postulated high pressure failure of PWR reactor vessel has been investigated by experimental works and one-dimensional computer modeling to find the relation between the fraction of melt simulant retained in the cavity and the reactor vessel initial conditions as well as to examine the hydrodynamic processes in a reactor cavity geometry. Simulated HPME experiments have been performed with two small-scale (1/25-th and 1/41-st) transparent reactor cavity models of the Young-Gwang unit 1 and 2. Wood's metal and water have been used as melt sumulants while high pressure nitrogen and carbon dioxide have been used as driver gases to simulate the blowdown steam and gas from the breach of the reactor pressure vessel. The high speed movies of the transient tests showed that no fraction of the melt simulant exits the cavity model via the vertical cavity tunnel under its own momentum, and that the discharged simulant from the pressure vessel exits the reactor cavity model during the gas blowdown. The principal removal mechanism seemed to be a combined mechanism of film entrainment and particle levitation due to the driving force of the blowdown gas. Experimental data for the fraction of melt simulant retained in the cavity model (Y f ) during a postulated scenario of the HPME from PWR pressure vessels have been obtained as a function of various test parameters. These data have been used to develop a correlation for Y f that fits all the data (a total of 313 data points) within the standard deviation of 0.054 by means of dimensional analysis and nonlinear least squares optimization technique. The basic effects of important parameters used to describe the HPME accident sequence on the Y f are determined based on the correlation obtained here and

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

The impact of vessel speed reduction on port accidents.

Science.gov (United States)

Chang, Young-Tae; Park, Hyosoo

2016-03-19

Reduced-speed zones (RSZs) have been designated across the world to control emissions from ships and prevent mammal strikes. While some studies have examined the effectiveness of speed reduction on emissions and mammal preservation, few have analyzed the effects of reduced ship speed on vessel safety. Those few studies have not yet measured the relationship between vessel speed and accidents by using real accident data. To fill this gap in the literature, this study estimates the impact of vessel speed reduction on vessel damages, casualties and frequency of vessel accidents. Accidents in RSZ ports were compared to non-RSZ ports by using U.S. Coast Guard data to capture the speed reduction effects. The results show that speed reduction influenced accident frequency as a result of two factors, the fuel price and the RSZ designation. Every $10 increase in the fuel price led to a 10.3% decrease in the number of accidents, and the RSZ designation reduced vessel accidents by 47.9%. However, the results do not clarify the exact impact of speed reduction on accident casualty. Copyright © 2016 Elsevier Ltd. All rights reserved.

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.

75 FR 54025 - Vessel and Facility Response Plans for Oil: 2003 Removal Equipment Requirements and Alternative...

Science.gov (United States)

2010-09-03

... responders for each vessel or facility with appropriate equipment and resources located in each zone of operation; specific lists of equipment that the resource providers will make available in case of an...

Comparative Study of Retinal Vessel Segmentation Based on Global Thresholding Techniques

Directory of Open Access Journals (Sweden)

Temitope Mapayi

2015-01-01

Full Text Available Due to noise from uneven contrast and illumination during acquisition process of retinal fundus images, the use of efficient preprocessing techniques is highly desirable to produce good retinal vessel segmentation results. This paper develops and compares the performance of different vessel segmentation techniques based on global thresholding using phase congruency and contrast limited adaptive histogram equalization (CLAHE for the preprocessing of the retinal images. The results obtained show that the combination of preprocessing technique, global thresholding, and postprocessing techniques must be carefully chosen to achieve a good segmentation performance.

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

Science.gov (United States)

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

2016-02-01

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

Green vessel scheduling in liner shipping: Modeling carbon dioxide emission costs in sea and at ports of call

Directory of Open Access Journals (Sweden)

Maxim A. Dulebenets

2018-03-01

Full Text Available Considering a substantial increase in volumes of the international seaborne trade and drastic climate changes due to carbon dioxide emissions, liner shipping companies have to improve planning of their vessel schedules and improve energy efficiency. This paper presents a novel mixed integer non-linear mathematical model for the green vessel scheduling problem, which directly accounts for the carbon dioxide emission costs in sea and at ports of call. The original non-linear model is linearized and then solved using CPLEX. A set of numerical experiments are conducted for a real-life liner shipping route to reveal managerial insights that can be of importance to liner shipping companies. Results indicate that the proposed mathematical model can serve as an efficient planning tool for liner shipping companies and may assist with evaluation of various carbon dioxide taxation schemes. Increasing carbon dioxide tax may substantially change the design of vessel schedules, incur additional route service costs, and improve the environmental sustainability. However, the effects from increasing carbon dioxide tax on the marine container terminal operations are found to be very limited.

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

Reactor vessel assessment and the development of a reactor vessel life extension program for Calvert Cliffs Units One and Two

International Nuclear Information System (INIS)

Montgomery, B.; Hijeck, P.J.

1988-01-01

A study has been undertaken to provide a general assessment of the life extension capabilities for the Calvert Cliffs Units One and Two reactor pressure vessels. The purpose of the study is to assess the general life extension capabilities for the Calvert Cliffs reactor pressure vessels based upon an extension and variation of the Surry pilot plant life extension study. This assessment provided a detailed reactor vessel surveillance program for plant life extension along with a hierarchy of specific tasks necessary for attaining maximum useful life. The assessment identified a number of critical issues which may impact life attainment and extension along with potential solutions to address these issues to ensure the life extension option is not precluded

Moessbauer study of miniature vessels from the Sican burial site of Huaca Loro

Energy Technology Data Exchange (ETDEWEB)

Cervantes, G., E-mail: gabriela.cervantes@pucp.pe [Pontificia Universidad Catolica del Peru (Peru); Shimada, I. [Southern Illinois University (United States); Haeusler, W.; Wagner, U.; Wagner, F. E. [Technische Universitaet Muenchen, Physik-Department E15 (Germany)

2011-11-15

Small handmade ceramic vessels are often found in large quantities in late pre-hispanic elite tombs on the north coast of Peru. These miniature vessels are of poor artistic and technologic qualities compared with other offerings found in tombs and have not received much attention. Our hypothesis is that these miniatures were made hastily by non-elite funeral attendants, which resulted in the observed formal and material variability. Here we report on the results of a systematic archaeological analysis of the shape and of archaeometric analyses by Moessbauer spectroscopy and XRD of the firing conditions of such vessels that were conducted to test the above hypothesis.

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

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)

Demonstration tests for manufacturing the ITER vacuum vessel

International Nuclear Information System (INIS)

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

2007-01-01

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

Demonstration tests for manufacturing the ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

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

2007-10-15

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

Quantification of tumor vessels in glioblastoma patients using time-of-flight angiography at 7 Tesla: a feasibility study.

Directory of Open Access Journals (Sweden)

Alexander Radbruch

Full Text Available PURPOSE: To analyze if tumor vessels can be visualized, segmented and quantified in glioblastoma patients with time of flight (ToF angiography at 7 Tesla and multiscale vessel enhancement filtering. MATERIALS AND METHODS: Twelve patients with newly diagnosed glioblastoma were examined with ToF angiography (TR = 15 ms, TE = 4.8 ms, flip angle = 15°, FOV = 160 × 210 mm(2, voxel size: 0.31 × 0.31 × 0.40 mm(3 on a whole-body 7 T MR system. A volume of interest (VOI was placed within the border of the contrast enhancing part on T1-weighted images of the glioblastoma and a reference VOI was placed in the non-affected contralateral white matter. Automated segmentation and quantification of vessels within the two VOIs was achieved using multiscale vessel enhancement filtering in ImageJ. RESULTS: Tumor vessels were clearly visible in all patients. When comparing tumor and the reference VOI, total vessel surface (45.3 ± 13.9 mm(2 vs. 29.0 ± 21.0 mm(2 (p<0.035 and number of branches (3.5 ± 1.8 vs. 1.0 ± 0.6 (p<0.001 per cubic centimeter were significantly higher, while mean vessel branch length was significantly lower (3.8 ± 1.5 mm vs 7.2 ± 2.8 mm (p<0.001 in the tumor. DISCUSSION: ToF angiography at 7-Tesla MRI enables characterization and quantification of the internal vascular morphology of glioblastoma and may be used for the evaluation of therapy response within future studies.

A study on probabilistic fracture mechanics for nuclear pressure vessels and piping

International Nuclear Information System (INIS)

Yagawa, Genki; Yoshimura, Shinobu

1997-01-01

This paper describes some recent research activities on probabilistic fracture mechanics (PFM) for nuclear pressure vessels and piping (PV and P) performed by the RC111 research committee of the Japan Society of Mechanical Engineers (JSME) under a subcontract of the Japan Atomic Energy Research Institute (JAERI). To establish standard procedures for evaluating failure probabilities of nuclear PV and P, we have set up the following three kinds of PFM round-robin problems on: (a) primary piping under normal operating conditions, (b) aged reactor pressure vessel (RPV) under normal and upset operating conditions, and (c) aged RPV under pressurised thermal shock (PTS) events. The basic problems of the last one are chosen from some US benchmark problems such as EPRI (Electric Power Research Institute) and US NRC (Nuclear Regulatory Commission) joint PTS benchmark problems. This paper summarizes some sensitivity studies on the three kinds of problems mainly varying material properties such as flow stress, fracture toughness, fatigue crack growth rate, Cu content. Employed in this study are the PFM computer codes developed in Japan and USA. Failure probabilities of nuclear PV and P are quantitatively discussed in detail. (author)

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

Cholinergic innervation of human mesenteric lymphatic vessels.

Science.gov (United States)

D'Andrea, V; Bianchi, E; Taurone, S; Mignini, F; Cavallotti, C; Artico, M

2013-11-01

The cholinergic neurotransmission within the human mesenteric lymphatic vessels has been poorly studied. Therefore, our aim is to analyse the cholinergic nerve fibres of lymphatic vessels using the traditional enzymatic techniques of staining, plus the biochemical modifications of acetylcholinesterase (AChE) activity. Specimens obtained from human mesenteric lymphatic vessels were subjected to the following experimental procedures: 1) drawing, cutting and staining of tissues; 2) staining of total nerve fibres; 3) enzymatic staining of cholinergic nerve fibres; 4) homogenisation of tissues; 5) biochemical amount of proteins; 6) biochemical amount of AChE activity; 6) quantitative analysis of images; 7) statistical analysis of data. The mesenteric lymphatic vessels show many AChE positive nerve fibres around their wall with an almost plexiform distribution. The incubation time was performed at 1 h (partial activity) and 6 h (total activity). Moreover, biochemical dosage of the same enzymatic activity confirms the results obtained with morphological methods. The homogenates of the studied tissues contain strong AChE activity. In our study, the lymphatic vessels appeared to contain few cholinergic nerve fibres. Therefore, it is expected that perivascular nerve stimulation stimulates cholinergic nerves innervating the mesenteric arteries to release the neurotransmitter AChE, which activates muscarinic or nicotinic receptors to modulate adrenergic neurotransmission. These results strongly suggest, that perivascular cholinergic nerves have little or no effect on the adrenergic nerve function in mesenteric arteries. The cholinergic nerves innervating mesenteric arteries do not mediate direct vascular responses.

Vessel size measurements in angiograms: A comparison of techniques

International Nuclear Information System (INIS)

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

2002-01-01

As interventional procedures become more complicated, the need for accurate quantitative vascular information increases. In response to this need, many commercial vendors provide techniques for measurement of vessel sizes, usually based on derivative techniques. In this study, we investigate the accuracy of several techniques used in the measurement of vessel size. Simulated images of vessels having circular cross sections were generated and convolved with various focal spot distributions taking into account the magnification. These vessel images were then convolved with Gaussian image detector line spread functions (LSFs). Additionally, images of a phantom containing vessels with a range of diameters were acquired for the 4.5'', 6'', 9'', and 12'' modes of an image intensifier-TV (II-TV) system. Vessel sizes in the images were determined using a first-derivative technique, a second-derivative technique, a linear combination of these two measured sizes, a thresholding technique, a densitometric technique, and a model-based technique. For the same focal spot size, the shape of the focal spot distribution does not affect measured vessel sizes except at large magnifications. For vessels with diameters larger than the full-width-at-half-maximum (FWHM) of the LSF, accurate vessel sizes (errors ∼0.1 mm) could be obtained by using an average of sizes determined by the first and second derivatives. For vessels with diameters smaller than the FWHM of the LSF, the densitometric and model-based techniques can provide accurate vessel sizes when these techniques are properly calibrated

Experimental study of the effect of neutron radiation on pressurised water reactor vessel steel resilience - First part

International Nuclear Information System (INIS)

Verdeau, Jean-Jacques

1969-12-01

After having outlined the importance of the embrittlement of vessel steels by neutrons during the exploitation of pressurised water reactors, the author reports a set of tests which aimed at determining the effect of neutron irradiation on vessel steel resilience for operated, under construction or projected pressurized water reactors. He also tries to highlight the influence of irradiation temperature and of initial thermal treatments, and to look for a restoration thermal treatment of neutron-induced damages which could be applied to the considered vessels. Tests were performed on V Charpy resilience samples. Some samples have been irradiated by the Pile Department of the Grenoble CEN and then broken by the Laboratory of very high activity, whereas other samples have been irradiated in a prototype vessel and broken by a Cadarache department. The author presents characteristics of the studied steels (chemical compositions, thermal treatments), describes sample irradiation conditions, and the method of assessment of the transition temperature after irradiation, presents experimental results, discusses their interpretation, and presents future tests to be performed [fr

MRP-227 Reactor vessel internals inspection planning and initial results at the Oconee nuclear station unit 2

International Nuclear Information System (INIS)

Davidsaver, S.B.; Fyfitch, S.; Whitaker, D.E.; Doss, R.L.

2015-01-01

The U.S. PWR industry has pro-actively developed generic inspection requirements and standards for reactor vessel (RV) internals. The Electric Power Research Institute (EPRI) Pressurized Water Reactor (PWR) Materials Reliability Program (MRP) has issued MRP-227-A and MRP-228 with mandatory and needed requirements based on the Nuclear Energy Institute (NEI) document NEI 03-08. The inspection and evaluation guidelines contained in MRP-227-A consider eight age-related degradation mechanisms: stress corrosion cracking (SCC), irradiation-assisted stress corrosion cracking (IASCC), wear, fatigue, thermal aging embrittlement, irradiation embrittlement, void swelling and irradiation growth, and thermal and irradiation-enhanced stress relaxation or irradiation-enhanced creep. This paper will discuss the decision planning efforts required for implementing the MRP-227-A and MRP-228 requirements and the results of these initial inspections at the Oconee Nuclear power station (ONS) units. Duke Energy and AREVA overcame a significant technology and NDE challenge by successfully completing the first-of-a-kind MRP-227-A scope requirements at ONS-1 in one outage below the estimated dose and with zero safety issues or events. This performance was repeated at ONS-2 a year later. The remote NDE tooling and processes developed to examine the MRP-227-A scope for ONS-1 and ONS-2 are transferable to other PWRs

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.

Resolution of the reactor vessel materials toughness safety issue; Task Action Plan A-11; Appendices C-K

International Nuclear Information System (INIS)

Johnson, R.E.

1981-09-01

The central problem in the Unresolved Safety Issue A-11, 'Reactor Vessel Materials Toughness,' was to provide guidance in performing analyses for reactor pressure vessels (RPVs) which fail to meet the toughness requirements during service life as a result of neutron radiation embrittlement. A technical team of recognized experts was organized to assist the NRC staff in addressing the problem. Using the foundation of the tearing modulus concept, which has been developed under earlier NRC sponsorship, relationships were obtained which provided approximate solutions to the RPV fracture problem with assumed beltline region flaws. Volume I of this report is a brief presentation of the problem and the results; Volume II provides the detailed technical foundations

Running Safety of Trains under Vessel-Bridge Collision

Directory of Open Access Journals (Sweden)

Yongle Li

2015-01-01

Full Text Available To optimize the sensor placement of the health monitoring system, the dynamic behavior of the train-bridge system subjected to vessel-collision should be studied in detail firstly. This study thus focuses on the characteristics of a train-bridge system under vessel-bridge collision. The process of the vessel-bridge collision is simulated numerically with a reliable finite element model (FEM. The dynamic responses of a single car and a train crossing a cable-stayed bridge are calculated. It is shown that the collision causes significant increase of the train’s lateral acceleration, lateral wheelset force, wheel unloading rate, and derailment coefficient. The effect of the collision on the train’s vertical acceleration is much smaller. In addition, parametric studies with various train’s positions, ship tonnage, and train speed are performed. If the train is closer to the vessel-bridge collision position or the ship tonnage is larger, the train will be more dangerous. There is a relatively high probability of running danger at a low speed, resulting from longer stay of the train on the bridge. The train’s position, the ship tonnage, and the train speed must be considered when determining the most adverse conditions for the trains running on bridges under vessel-bridge collision.

Multi-detector spiral CT study of the relationships between pulmonary ground-glass nodules and blood vessels

International Nuclear Information System (INIS)

Gao, Feng; Li, Ming; Ge, Xiaojun; Ren, Qingguo; Hua, Yanqing; Zheng, Xiangpeng; Chen, Yan; Lv, Fangzhen

2013-01-01

To investigate the relationships between pulmonary ground-glass nodules (GGN) and blood vessels and their diagnostic values in differentiating GGNs. Multi-detector spiral CT imaging of 108 GGNs was retrospectively reviewed. The spatial relationships between GGNs and supplying blood vessels were categorized into four types: I, vessels passing by GGNs; II, intact vessels passing through GGNs; III, distorted, dilated or tortuous vessels seen within GGNs; IV, more complicated vasculature other than described above. Relationship types were correlated to pathologic and/or clinical findings of GGNs. Of 108 GGNs, 10 were benign, 24 preinvasive nodules and 74 adenocarcinomas that were pathologically proven. Types I, II, III and IV vascular relationships were observed in 9, 58, 21 and 20 GGNs, respectively. Type II relationship was the dominating relationship for each GGN group, but significant differences were shown among them. Correlation analysis showed strong correlation between invasive adenocarcinoma and type III and IV relationships. Subgroup analysis indicated that type III was more commonly seen in IAC with comparison to type IV more likely seen in MIA. Different GGNs have different relationships with vessels. Understanding and recognising characteristic GGN-vessel relationships may help identify which GGNs are more likely to be malignant. (orig.)

Optimization of Helium Vessel Design for ILC Cavities

Energy Technology Data Exchange (ETDEWEB)

Fratangelo, Enrico [Univ. of Pisa (Italy)

2009-01-01

The ILC (International Linear Collider) is a proposed new major particle accelerator. It consists of two 20 km long linear accelerators colliding electrons and positrons at an energy exceeding 500 GeV, Achieving this collision energy while keeping reasonable accelerator dimensions requires the use of high electric field superconducting cavities as the main acceleration element. These cavities are operated at l.3 GHz inside an appropriate container (He vessel) at temperatures as low as 1.4 K using superfluid Helium as the refrigerating medium. The purpose of this thesis, in the context of the ILC R&D activities currently in progress at Fermilab (Fermi National Accelerator Laboratory), is the mechanical study of an ILC superconducting cavity and Helium vessel prototype. The main goals of these studies are the determination of the limiting working conditions of the whole He vessel assembly, the simulation of the manufacturing process of the cavity end-caps and the assessment of the Helium vessel's efficiency. In addition this thesis studies the requirements to certify the compliance with the ASME Code of the whole cavity/vessel assembly. Several Finite Elements Analyses were performed by the candidate himself in order to perform the studies listed above and described in detail in Chapters 4 through 8. ln particular the candidate has developed an improved procedure to obtain more accurate results with lower computational times. These procedures will be accurately described in the following chapters. After an introduction that briefly describes the Fennilab and in particular the Technical Division (where all the activities concerning with this thesis were developed), the first part of this thesis (Chapters 2 and 3) explains some of the main aspects of modem particle accelerators. Moreover it describes the most important particle accelerators working at the moment and the basic features of the ILC project. Chapter 4 describes all the activities that were done to

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

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

Benchmark study of shear buckling of a cylindrical vessel

International Nuclear Information System (INIS)

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

1986-01-01

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

Applicability of newly developed 610MPa class heavy thickness high strength steel to boiler pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Katayama, Norihiko; Kaihara, Shoichiro; Ishii, Jun [Ishikawajima-Harima Heavy Industries Corp., Yokohama (Japan); Kajigaya, Ichiro [Ishikawajima-Harima Heavy Industries Corp., Tokyo (Japan); Totsuka, Takehiro; Miyazaki, Takashi [Ishikawajima-Harima Heavy Industries Corp., Aioi (Japan)

1995-11-01

Construction of a 350 MW Class PFBC (Pressurized Fluidized Bed Combustion) boiler plant is under planning in Japan. Design temperature and pressure of the vessel are maximum 350 C and 1.69 MPa, respectively. As the plate thickness of the vessel exceeds over 100 mm, high strength steel plate of good weldability and less susceptible to reheat cracking was required and developed. The steel was aimed to satisfy the tensile strength over 610 MPa at 350 C after postweld heat treatment (PWHT), with good notch toughness. The authors investigated the welding performances of the newly developed steel by using 150 mm-thick plate welded by pulsed-MAG and SAW methods. It was confirmed that the newly developed steel and its welds possess sufficient strength and toughness after PWHT, and applicable to the actual pressure vessel.

Does bipolar electrocoagulation time affect vessel weld strength?

Science.gov (United States)

Harrison, J D; Morris, D L

1991-01-01

The value of the bipolar electrocoagulator in the haemostasis of bleeding ulcers is controversial. We have therefore investigated the effect of different coagulation times on vessel weld strength achieved by the bipolar device. Welds were then made in vessels of known diameter using a standard 10F endoscopic haemostatic probe at coagulation times of two and 20 seconds. The intravascular temperature achieved at each time was measured. Vessel weld strength achieved by bipolar electrocoagulation was much greater at 20 seconds (approximately twice that at two seconds) and was highly significantly greater at all vessel diameters. There was a gradual reduction in weld strength with increasing vessel diameter, an effect that was seen for both two and 20 seconds of electrocoagulation. Intravascular temperature was significantly higher at 20 seconds than at two seconds. We conclude that vessel weld strength is related to coagulation time and that any future studies comparing the bipolar electrocoagulator with other haemostatic devices should use longer periods of bipolar electrocoagulation and record the coagulation time in order to optimise the clinical value of the device. PMID:1864540

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

Distribution of normal superficial ocular vessels in digital images.

Science.gov (United States)

Banaee, Touka; Ehsaei, Asieh; Pourreza, Hamidreza; Khajedaluee, Mohammad; Abrishami, Mojtaba; Basiri, Mohsen; Daneshvar Kakhki, Ramin; Pourreza, Reza

2014-02-01

To investigate the distribution of different-sized vessels in the digital images of the ocular surface, an endeavor which may provide useful information for future studies. This study included 295 healthy individuals. From each participant, four digital photographs of the superior and inferior conjunctivae of both eyes, with a fixed succession of photography (right upper, right lower, left upper, left lower), were taken with a slit lamp mounted camera. Photographs were then analyzed by a previously described algorithm for vessel detection in the digital images. The area (of the image) occupied by vessels (AOV) of different sizes was measured. Height, weight, fasting blood sugar (FBS) and hemoglobin levels were also measured and the relationship between these parameters and the AOV was investigated. These findings indicated a statistically significant difference in the distribution of the AOV among the four conjunctival areas. No significant correlations were noted between the AOV of each conjunctival area and the different demographic and biometric factors. Medium-sized vessels were the most abundant vessels in the photographs of the four investigated conjunctival areas. The AOV of the different sizes of vessels follows a normal distribution curve in the four areas of the conjunctiva. The distribution of the vessels in successive photographs changes in a specific manner, with the mean AOV becoming larger as the photos were taken from the right upper to the left lower area. The AOV of vessel sizes has a normal distribution curve and medium-sized vessels occupy the largest area of the photograph. Copyright © 2013 British Contact Lens Association. Published by Elsevier Ltd. All rights reserved.

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)

Gammatography of thick lead vessels

International Nuclear Information System (INIS)

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

1979-01-01

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

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

Science.gov (United States)

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

2017-10-01

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

Opinion on the demonstration of the 900 MWe reactor vessels in-service behaviour after their third decennial inspection

International Nuclear Information System (INIS)

2010-01-01

In this report, an expert group comments and assesses how sufficient are the demonstration and the actions performed by EDF to justify the in-service behaviour of nuclear reactor vessels. More precisely, it comments and discusses the different steps of the EDF demonstration: follow-up of the fluence received by the vessels, identification of the most severe transients and thermodynamic calculations, behaviour of irradiated materials, mechanical analysis, in-service control and follow-up plan, ageing management. Recommendations are then formulated

Guide to the periodic inspection of nuclear reactor steel pressure vessels

International Nuclear Information System (INIS)

1969-01-01

This Guide is intended to provide general information and guidance to reactor owners or operators, inspection authorities, certifying authorities or regulatory bodies who are responsible for establishing inspection procedures for specific reactors or reactor types, and for the preparation of national codes or standards. The recommendations of the Guide apply primarily to water-cooled steel reactor vessels which are at a sufficiently early stage of design so that recommendations to provide accessibility for inspection can be incorporated into the early stages of design and inspection planning. However, much of the contents of the Guide are also applicable in part to vessels for other reactor types, such as gas-cooled, pressure-tube, or liquid-metal-cooled reactors, and also to some existing water-cooled reactors and reactors which are in advanced stage of design or construction. 46 refs, figs, 1 tab

Retinal Vessels Segmentation Techniques and Algorithms: A Survey

Directory of Open Access Journals (Sweden)

Jasem Almotiri

2018-01-01

Full Text Available Retinal vessels identification and localization aim to separate the different retinal vasculature structure tissues, either wide or narrow ones, from the fundus image background and other retinal anatomical structures such as optic disc, macula, and abnormal lesions. Retinal vessels identification studies are attracting more and more attention in recent years due to non-invasive fundus imaging and the crucial information contained in vasculature structure which is helpful for the detection and diagnosis of a variety of retinal pathologies included but not limited to: Diabetic Retinopathy (DR, glaucoma, hypertension, and Age-related Macular Degeneration (AMD. With the development of almost two decades, the innovative approaches applying computer-aided techniques for segmenting retinal vessels are becoming more and more crucial and coming closer to routine clinical applications. The purpose of this paper is to provide a comprehensive overview for retinal vessels segmentation techniques. Firstly, a brief introduction to retinal fundus photography and imaging modalities of retinal images is given. Then, the preprocessing operations and the state of the art methods of retinal vessels identification are introduced. Moreover, the evaluation and validation of the results of retinal vessels segmentation are discussed. Finally, an objective assessment is presented and future developments and trends are addressed for retinal vessels identification techniques.

Limiting Factors for External Reactor Vessel Cooling

International Nuclear Information System (INIS)

Cheung, F.B.

2005-01-01

different angular locations of the vessel outer surface could be enhanced by ∼1.2 to 2 times the CHF compared with a plain vessel without coatings. The CHF enhancement could be attributed to the structure of the porous coating itself and the capillary action it induced. The matrix of cavities and voids within the coating effectively trap vapor, which serve as active nucleation sites. These sites in turn are fed with liquid flowing through the interconnected channels. The pores on the surface of the porous coating serve as flow inlets for liquid supply to the heating surface, leading to appreciable enhancement in downward-facing boiling heat transfer and the local CHF limits. Results of the present study suggest that by utilizing an enhanced vessel/insulation design with vessel coating, it is possible to significantly enhance the CLSV and the CHF limits as well as minimize the two-phase flow instability problems, thus substantially increasing the margin for IVR

Low lymphatic vessel density associates with chronic rhinosinusitis with nasal polyps.

Science.gov (United States)

Luukkainen, A; Seppälä, M; Renkonen, J; Renkonen, R; Hagstrő M, J; Huhtala, H; Rautiainen, M; Myller, J; Paavonen, T; Ranta, A; Torkkeli, T; Toppila-Salmi, S

2017-06-01

Chronic rhinosinusitis with and without nasal polyps (CRSwNP and CRSsNP) and antrochoanal polyps (ACP) are different upper airway inflammation phenotypes with different pathomechanisms. In order to understand the development of tissue edema, the present study aimed to evaluate lymphatic vessel density in CRSsNP, CRSwNP and ACP. 120 retrospective nasal and maxillary sinus specimens were stained immunohistochemically with a von Willebrand factor polyclonal antibody recognizing vascular and lymphatic endothelium, and with a podoplanin monoclonal antibody recognizing lymphatic endothelium. Vessels were studied by microscopy in a blinded fashion, and the vessel density and the relative density of lymphatic vessels were calculated. Patient characteristic factors and follow-up data of in average 9 years were collected from patient records. In the nasal cavity, the low absolute and relative density of vessels and of lymphatic vessels was associated with CRSwNP and ACP tissues compared to control inferior turbinate. This was observed also in the inflammatory hotspot area. In the maxillary sinus, lower absolute and relative density of lymphatic vessels associated with the CRSwNP phenotype. High lymphatic vessel density in polyp tissue associated with the need for revision CRS-surgery. As a conclusion, low density of lymphatic vessels distinguished patients with CRSwNP not only in the hotspot area of polyp tissue, but also in maxillary sinus mucosa. Yet, higher lymphatic vessel density seems to associate with polyp recurrence. Further studies are still needed to explore if formation of nasal polyps could be diminished by intranasal therapeutics affecting lymphangiogenesis.

Comparison of the number of gingival blood vessels between type 2 diabetes mellitus and chronic periodontitis patients: An immunohistological study

Directory of Open Access Journals (Sweden)

Gautami Subhadra Penmetsa

2015-01-01

Full Text Available Background: The relationship between diabetes and periodontitis has been studied for more than 50 years and is generally agreed that the periodontal disease is more prevalent in diabetic patients compared to nondiabetics. Vascular changes like increased thickness of basement membrane in small vessels has been reported in diabetic patients, but the quantity of blood vessels in gingiva of diabetic patients has not been discussed much. The aim of this study was to compare the number of blood vessels in gingiva between chronic periodontitis (CP patients, CP with diabetes (type 2, and normal healthy gingiva. Materials and Methods: The study included 75 patients, divided into three groups of 25 patients each-Group I with healthy periodontium (HP, Group II with CP, and Group III with CP with diabetes mellitus (CPDM.Gingival biopsies were obtained from the subjects undergoing crown lengthening procedure for Group I, and in patients with CP and in CPDM biopsies were collected from teeth undergoing extraction. Sections were prepared for immune histochemical staining with CD34. Results: Difference was observed in the average number of blood vessels when compared between HP, CP, and CPDM groups. Statistical significant difference was observed when the HP and CP groups and HP and CPDM groups were compared. Conclusion: The results of the study indicated that the number of blood vessels in gingival connective tissue is significantly higher in CP and CPDM patients.

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

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

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.)

A computational study for investigating acoustic streaming and tissue heating during high intensity focused ultrasound through blood vessel with an obstacle

Science.gov (United States)

Parvin, Salma; Sultana, Aysha

2017-06-01

The influence of High Intensity Focused Ultrasound (HIFU) on the obstacle through blood vessel is studied numerically. A three-dimensional acoustics-thermal-fluid coupling model is employed to compute the temperature field around the obstacle through blood vessel. The model construction is based on the linear Westervelt and conjugate heat transfer equations for the obstacle through blood vessel. The system of equations is solved using Finite Element Method (FEM). We found from this three-dimensional numerical study that the rate of heat transfer is increasing from the obstacle and both the convective cooling and acoustic streaming can considerably change the temperature field.

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

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.)

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

Ultrasonographic Examination of Some Vessels in Dogs and the Characteristics of Blood Flow in These Vessels

Directory of Open Access Journals (Sweden)

Figurová M.

2017-12-01

Full Text Available The examination by Doppler ultrasonography provides haemodynamic information about blood flow velocity in a respective vessel. It specifies high- and lowresistance flow patterns. The aim of our study was to record the flow in a. carotis communis, a. femoralis and aa. renales in 16 adult clinically healthy dogs of small and medium size; characterize the types of vessels and also determine the pulsatility index (PI and the resistive index (RI of these vessels. The a. femoralis is a high-resistance vessel with a pronounced three-peak waveform. The aa. renales gives a typical picture of a low-resistance flow pattern. The characteristics of a. carotis communis involves different images of its branches a. carotis interna and a. carotis externa. In the investigated groups we observed a medium degree of pulsatility (atypical highresistance flow pattern with an absence of reverse flow. The mean measured values of indices for a. carotis communis were: left side PI 1.824 and RI 0.742; right side PI 1.891 and RI 0.746, and for aa. renales: PI 1.366 ± 0.04 and RI 0.684 ± 0.05.

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...

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.)

Design characteristics of pantograph type in vessel fuel handling system in SFR

International Nuclear Information System (INIS)

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

2012-01-01

The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied

Design characteristics of pantograph type in vessel fuel handling system in SFR

Energy Technology Data Exchange (ETDEWEB)

Kim, S. H.; Koo, G. H. [KAERI, Daejeon (Korea, Republic of)

2012-10-15

The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied.

75 FR 15391 - Approval and Promulgation of Air Quality Implementation Plans; Texas; Revision to Control...

Science.gov (United States)

2010-03-29

... to control volatile organic compound (VOC) emissions from storage tanks, transport vessels and marine... Promulgation of Air Quality Implementation Plans; Texas; Revision to Control Volatile Organic Compound.... Specifically, this revision subjects owners or operators of VOC storage tanks, transport vessels, and marine...

Trends in Tissue Engineering for Blood Vessels

Directory of Open Access Journals (Sweden)

Judee Grace Nemeno-Guanzon

2012-01-01

Full Text Available Over the years, cardiovascular diseases continue to increase and affect not only human health but also the economic stability worldwide. The advancement in tissue engineering is contributing a lot in dealing with this immediate need of alleviating human health. Blood vessel diseases are considered as major cardiovascular health problems. Although blood vessel transplantation is the most convenient treatment, it has been delimited due to scarcity of donors and the patient’s conditions. However, tissue-engineered blood vessels are promising alternatives as mode of treatment for blood vessel defects. The purpose of this paper is to show the importance of the advancement on biofabrication technology for treatment of soft tissue defects particularly for vascular tissues. This will also provide an overview and update on the current status of tissue reconstruction especially from autologous stem cells, scaffolds, and scaffold-free cellular transplantable constructs. The discussion of this paper will be focused on the historical view of cardiovascular tissue engineering and stem cell biology. The representative studies featured in this paper are limited within the last decade in order to trace the trend and evolution of techniques for blood vessel tissue engineering.

Study of radiation damage of steels for light water pressure vessels at UJV

International Nuclear Information System (INIS)

Vacek, N.; Stoces, B.

1980-01-01

Preoperational determination of radiation resistance of pressure vessel steels is performed at accelerated neutron exposure in a test or materials research reactor. The results obtained at accelerated and operating exposure are not fully identical and surveillance bodies are therefore used manufactured from the pressure vessel material. Currently, the following steels are used for the manufacture of light water reactor pressure vessels: Mn-Mo-Ni (ASTM-A533-B, ASTM-A508), Cr-Mo-V (15Kh2M1FA). At UJV Rez, for irradiation Chanca-M probes imported from France are used featuring electric temperature control. Almost identical radiation embrittlement was measured for all three steels after irradiation with a neutron fluence of 3x10 23 n.m -2 at a temperature of 290 degC. (H.S.)

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Tumors induce coordinate growth of artery, vein, and lymphatic vessel triads

International Nuclear Information System (INIS)

Ruddell, Alanna; Croft, Alexandra; Kelly-Spratt, Karen; Furuya, Momoko; Kemp, Christopher J

2014-01-01

Tumors drive blood vessel growth to obtain oxygen and nutrients to support tumor expansion, and they also can induce lymphatic vessel growth to facilitate fluid drainage and metastasis. These processes have generally been studied separately, so that it is not known how peritumoral blood and lymphatic vessels grow relative to each other. The murine B16-F10 melanoma and chemically-induced squamous cell carcinoma models were employed to analyze large red-colored vessels growing between flank tumors and draining lymph nodes. Immunostaining and microscopy in combination with dye injection studies were used to characterize these vessels. Each peritumoral red-colored vessel was found to consist of a triad of collecting lymphatic vessel, vein, and artery, that were all enlarged. Peritumoral veins and arteries were both functional, as detected by intravenous dye injection. The enlarged lymphatic vessels were functional in most mice by subcutaneous dye injection assay, however tumor growth sometimes blocked lymph drainage to regional lymph nodes. Large red-colored vessels also grew between benign papillomas or invasive squamous cell carcinomas and regional lymph nodes in chemical carcinogen-treated mice. Immunostaining of the red-colored vessels again identified the clustered growth of enlarged collecting lymphatics, veins, and arteries in the vicinity of these spontaneously arising tumors. Implanted and spontaneously arising tumors induce coordinate growth of blood and lymphatic vessel triads. Many of these vessel triads are enlarged over several cm distance between the tumor and regional lymph nodes. Lymphatic drainage was sometimes blocked in mice before lymph node metastasis was detected, suggesting that an unknown mechanism alters lymph drainage patterns before tumors reach draining lymph nodes

Tumors induce coordinate growth of artery, vein, and lymphatic vessel triads.

Science.gov (United States)

Ruddell, Alanna; Croft, Alexandra; Kelly-Spratt, Karen; Furuya, Momoko; Kemp, Christopher J

2014-05-21

Tumors drive blood vessel growth to obtain oxygen and nutrients to support tumor expansion, and they also can induce lymphatic vessel growth to facilitate fluid drainage and metastasis. These processes have generally been studied separately, so that it is not known how peritumoral blood and lymphatic vessels grow relative to each other. The murine B16-F10 melanoma and chemically-induced squamous cell carcinoma models were employed to analyze large red-colored vessels growing between flank tumors and draining lymph nodes. Immunostaining and microscopy in combination with dye injection studies were used to characterize these vessels. Each peritumoral red-colored vessel was found to consist of a triad of collecting lymphatic vessel, vein, and artery, that were all enlarged. Peritumoral veins and arteries were both functional, as detected by intravenous dye injection. The enlarged lymphatic vessels were functional in most mice by subcutaneous dye injection assay, however tumor growth sometimes blocked lymph drainage to regional lymph nodes. Large red-colored vessels also grew between benign papillomas or invasive squamous cell carcinomas and regional lymph nodes in chemical carcinogen-treated mice. Immunostaining of the red-colored vessels again identified the clustered growth of enlarged collecting lymphatics, veins, and arteries in the vicinity of these spontaneously arising tumors. Implanted and spontaneously arising tumors induce coordinate growth of blood and lymphatic vessel triads. Many of these vessel triads are enlarged over several cm distance between the tumor and regional lymph nodes. Lymphatic drainage was sometimes blocked in mice before lymph node metastasis was detected, suggesting that an unknown mechanism alters lymph drainage patterns before tumors reach draining lymph nodes.

An experimental study on coolability through the external reactor vessel cooling according to RPV insulation design

Energy Technology Data Exchange (ETDEWEB)

Kang, Kyoung Ho; Koo, Kil Mo; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik

2004-01-01

LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the water accessibility and coolability in case of the external reactor vessel cooling. Alumina iron thermite melt was used as corium stimulant. And the hemispherical test vessel is linearly scaled-down of RPV lower plenum. 4 tests have been performed varying the melt composition and the configuration of the insulation system. Due to the limited steam venting capacity through the insulation, steam binding occurred inside the annulus in the LAVA- ERVC-1, 2 tests which were performed for simulating the KSNP insulation design. This steam binding brought about incident heat up of the vessel outer surface at the upper part in the LAVA-ERVC-1, 2 tests. On the contrary, in the LAVA-ERVC-3, 4 tests which were performed for simulating the APR1400 insulation design, the temperatures of the vessel outer surface maintained near saturation temperature. Sufficient water ingression and steam venting through the insulation lead to effective cooldown of the vessel characterized by nucleate boiling in the LAVA-ERVC-3, 4 tests. From the LAVA-ERVC experimental results, it could be preliminarily concluded that if pertinent modification of the insulation design focused on the improvement of water ingression and steam venting should be preceded the possibility of in-vessel corium retention through the external vessel cooling could be considerably increased.

Confinement Vessel Assay System: Design and Implementation Report

International Nuclear Information System (INIS)

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

2012-01-01

Los Alamos National Laboratory has a number of spherical confinement vessels remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1- to 2-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. We have developed a neutron assay system for the purposes of Materials Control and Accountability (MC and A) measurements of the vessel prior to and after cleanout. We present our approach to confronting the challenges in designing, building, and testing such a system. The system was designed to meet a set of functional and operational requirements. A Monte Carlo model was developed to aid in optimizing the detector design as well as to predict the systematic uncertainty associated with confinement vessel measurements. Initial testing was performed to optimize and determine various measurement parameters, and then the system was characterized using 252 Cf placed a various locations throughout the measurement system. Measurements were also performed with a 252 Cf source placed inside of small steel and HDPE shells to study the effect of moderation. These measurements compare favorably with their MCNPX model equivalent, making us confident that we can rely on the Monte Carlo simulation to predict the systematic uncertainty due to variations in response to material that may be localized at different points within a vessel.

RAPID COMMUNICATION: Magnetic resonance imaging inside metallic vessels

Science.gov (United States)

Han, Hui; Balcom, Bruce J.

2010-10-01

We introduce magnetic resonance imaging (MRI) measurements inside metallic vessels. Until now, MRI has been unusable inside metallic vessels because of eddy currents in the walls. We have solved the problem and generated high quality images by employing a magnetic field gradient monitoring method. The ability to image within metal enclosures and structures means many new samples and systems are now amenable to MRI. Most importantly this study will form the basis of new MRI-compatible metallic pressure vessels, which will permit MRI of macroscopic systems at high pressure.

Prestressed concrete pressure vessels for boiling water reactors

International Nuclear Information System (INIS)

Menon, S.

1979-12-01

Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

Analisa Greenwater Akibat Gerakan Offshore Security Vessel

Directory of Open Access Journals (Sweden)

Maulidya Octaviani Bustamin

2012-09-01

Full Text Available Analisa  Tugas  Akhir  ini,  terdiri  atas  beberapa  tahapan.  Yang pertama yaitu perancangan struktur Offshore Security Vessel (OSV dengan bantuan software MAXSURF guna mendapatkan Lines Plan. Offset data yang diperoleh digunakan dalam pemodelan menggunakan MOSES,  kemudian  dilakukan  analisa  gerak  OSV  dalam  gelombang  regular  dan dinyatakan dalam grafik RAO. Analisa gerak relatif vertikal  haluan dihitung dari RAO gerakan, dan kemudian melakukan evaluasi perilaku di gelombang acak dengan analisis spektra gelombang. Dari analisa spektra didapatkan parameter greenwater sehingga dapat dihitung peluang, intensitas dan tekanan greenwater. Dari hasil analisa diperoleh RAO gerak vertikal Offshore Security Vessel (OSV pada  gelombang  reguler yang dipengaruhi  oleh  kecepatan,  kondisi  muatan  dan arah gelombang. Peluang terjadinya greenwater terbesar terjadi pada sudut datang gelombang following sea (0o dimana harga terbesar terjadi pada ω = 0.2 rad/sec dengan periode 29 detik mencapai 0.477. Intensitas greenwater terbesar terjadi pada saat sudut datang gelombang following sea (0o adalah sebanyak 59.265 per jam dan 0.378 per detik. Tekanan greenwater terbesar terjadi pada saat sudut datang gelombang head sea (180o sebesar 1678x10-6 MPa. Dengan nilai tersebut, deck mampu menahan beban akibat tekanan greenwater.

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...

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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.

Study of the efficiency of the anti-convective thermal barrier of the Super-Phenix vessels inter space

International Nuclear Information System (INIS)

Durin, M.; Mejane, A.

1983-08-01

In the LMFBR Phenix reactor, the junction between the primary vessel and the roof slab is a region of large thermal gradients. In order to limit the gradient in the primary vessel, a thermal barrier has been installed between the primary and the safety vessel. The purpose of this barrier is to prevent the penetration of hot gas in the upper part of the vessels inter space. Experimental results have been obtained on a full scale model representing a 25 0 vessel sector of the reactor. Different geometrical configurations have been tested for a large range of boundary condition: - perfectly tight barrier - no thermal barrier; - simulation of leakages on the barrier [fr

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Advanced in-vessel retention design for next generation risk management

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-31

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

Advanced in-vessel retention design for next generation risk management

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

Surgical anatomy of penis in exstrophy-epispadias: a study of arrangement of fascial planes and superficial vessels of surgical significance.

Science.gov (United States)

Kureel, Shiv Narain; Gupta, Archika; Singh, Chandra Shekhar; Kumar, Manoj

2013-10-01

To study the anatomic arrangement of the fascial planes and superficial vessels in relationship to the laid-open urethral plate, glans, corpus spongiosum, and corpora cavernosa in the penis of patients with exstrophy or epispadias. Of 6 patients, 4 had classic exstrophy and 2 had incontinent epispadias. These patients had presented beyond adolescence without previous intervention and were selected for the present study. Using a 1.5-T magnetic resonance imaging scanner and compatible 3-in. surface coil, the epispadiac penises were studied using fast spin echo sequences and contrast-enhanced sequences. In 2 patients, angiography of the superficial vessels was also performed using multidetector row helical computed tomography. The imaging findings were also verified during the subsequent reconstructive surgery. A clear demarcation of the skin, dartos fascia, Buck's fascia, corpora cavernosa, corpus spongiosum, and the intraglanular planes were seen with the course of the blood vessels. The penile dartos received axial pattern vessels from the external pudendal vessels, with collateral branches from the dorsal penile artery as transverse branches at the shaft of the penis and preputial branches at the coronal sulcus. Buck's fascia sleeved the corpora cavernosa, enveloped the neurovascular bundle, and fused with the corpus spongiosum without crossing the midline. Intraglanular extension of Buck's fascia separated the intraglanular vascular arcade from the tip of the corpora. Parallel to the ventral midline, axial pattern vessels to the skin-dartos complex are present, with an additional supply to the prepuce from the terminal penile arteries. These findings can be used for designing the skin coverage. The subfascial plane between the tip of the corpora and the intraglanular vascular arcade and the plane of cleavage between the cavernosa-spongiosum interface can be used for efficient corporal urethral separation. Copyright © 2013 Elsevier Inc. All rights reserved.

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

NARCIS (Netherlands)

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

1997-01-01

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

33 CFR Appendix C to Part 155 - Training Elements for Oil Spill Response Plans

Science.gov (United States)

2010-07-01

.... 155, App. C Appendix C to Part 155—Training Elements for Oil Spill Response Plans 1. General 1.1The portion of the plan dealing with training is one of the key elements of a response plan. This concept is... included training as one of the sections required in a vessel or facility response plan. In reviewing...

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)

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

Fabrication progress of the ITER vacuum vessel sector in Korea

Energy Technology Data Exchange (ETDEWEB)

Kim, B.C., E-mail: bckim@nfri.re.kr [National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon (Korea, Republic of); Lee, Y.J.; Hong, K.H.; Sa, J.W.; Kim, H.S.; Park, C.K.; Ahn, H.J.; Bak, J.S.; Jung, K.J. [National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon (Korea, Republic of); Park, K.H.; Roh, B.R.; Kim, T.S.; Lee, J.S.; Jung, Y.H.; Sung, H.J.; Choi, S.Y.; Kim, H.G.; Kwon, I.K.; Kwon, T.H. [Hyundai Heavy Industries Co. Ltd., Dong-gu, Ulsan (Korea, Republic of)

2013-10-15

Highlights: ► Fabrication of ITER vacuum vessel sector full scale mock-up to develop fabrication procedures. ► The welding and nondestructive examination techniques conform to RCC-MR. ► The preparation of real manufacturing of ITER vacuum vessel sector. -- Abstract: As a participant of ITER project, ITER Korea has to supply two ITER vacuum vessel sectors (Sector no. 6, no. 1) of total nine ITER VV sectors. After the procurement arrangement with ITER Organization, ITER Korea made the contract with Hyundai Heavy Industries (HHI) for fabrication of two sectors. Then the start of the manufacturing design was initiated from January 2010. HHI made three real scale R and D mock-ups to verify the critical fabrication feasibility issues on electron beam welding, 3D forming, welding distortion and achievable tolerances. The documentation according to IO and the French nuclear safety regulation requirement, the qualification of welding and nondestructive examination procedures conform to RCC-MR 2007 were proceed in parallel. The mass production of raw material was done after receiving ANB (agreed notified body) verification of product/parts and shop qualification. The manufacturing drawing, manufacturing and inspection plan of VV sector with supporting fabrication procedures are also verified by ANB, accordingly the first cutting and forming of plates for VV sector fabrication started from February 2012. This paper reports the latest fabrication progress of ITER vacuum vessel Sector no. 6 that will be assembled as the first sector in the ITER pit. The overall fabrication route, R and D mock-up fabrication results with forming and welding distortion analysis, qualification status of welding and nondestructive examination (NDE) are also presented.

Summary of Reported Whale-Vessel Collisions in Alaskan Waters

Directory of Open Access Journals (Sweden)

Janet L. Neilson

2012-01-01

Full Text Available Here we summarize 108 reported whale-vessel collisions in Alaska from 1978–2011, of which 25 are known to have resulted in the whale's death. We found 89 definite and 19 possible/probable strikes based on standard criteria we created for this study. Most strikes involved humpback whales (86% with six other species documented. Small vessel strikes were most common (<15 m, 60%, but medium (15–79 m, 27% and large (≥80 m, 13% vessels also struck whales. Among the 25 mortalities, vessel length was known in seven cases (190–294 m and vessel speed was known in three cases (12–19 kn. In 36 cases, human injury or property damage resulted from the collision, and at least 15 people were thrown into the water. In 15 cases humpback whales struck anchored or drifting vessels, suggesting the whales did not detect the vessels. Documenting collisions in Alaska will remain challenging due to remoteness and resource limitations. For a better understanding of the factors contributing to lethal collisions, we recommend (1 systematic documentation of collisions, including vessel size and speed; (2 greater efforts to necropsy stranded whales; (3 using experienced teams focused on determining cause of death; (4 using standard criteria for validating collision reports, such as those presented in this paper.

Stowage Planning in Multiple Ports with Shifting Fee Minimization

Directory of Open Access Journals (Sweden)

E. Zhang

2018-01-01

Full Text Available This paper studies the problem of stowage planning within a vessel bay in a multiple port transportation route, aiming at minimizing the total container shifting fee. Since the access to containers is in the top-to-bottom order for each stack, reshuffle operations occur when a target container to be unloaded at its destination port is not stowed on the top of a stack at the time. Each container shift via a quay crane induces one unit of shifting fee that depends on the charge policy of the local container port. Previous studies assume that each container shift consumes a uniform cost in all ports and thus focus on minimizing the total number of shifts or the turnaround time of the vessel. Motivated by the observation that different ports are of nonuniform fee for each container shift, we propose a mixed integer programming (MIP model for the problem to produce an optimal stowage planning with minimum total shifting fee in this work. Moreover, as the considered problem is NP-hard due to the NP-hardness of its counterpart with uniform unit shifting fee, we propose an improved genetic algorithm to solve the problem. The efficiency of the proposed algorithm is demonstrated via numerical experiments.

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

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

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

International Nuclear Information System (INIS)

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

1990-01-01

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

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.)

[Modification of the Hungate vessel for cultivation of facultative and obligate anaerobic bacteria].

Science.gov (United States)

Chernyshenko, D V

2000-01-01

Modified Hungate vessel made of native penicillinum bottles and chemical vessels has been created and experimentally studied. The vessels can be used for cultivation of facultative and obligate anaerobe microorganisms on liquid and solid nutrient media. Locking devices of the vessel are described.

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

Experimental and numerical study on coupled motion responses of a floating crane vessel and a lifted subsea manifold in deep water

Directory of Open Access Journals (Sweden)

B.W. Nam

2017-09-01

Full Text Available The floating crane vessel in waves gives rise to the motion of the lifted object which is connected to the hoisting wire. The dynamic tension induced by the lifted object also affects the motion responses of the floating crane vessel in return. In this study, coupled motion responses of a floating crane vessel and a lifted subsea manifold during deep-water installation operations were investigated by both experiments and numerical calculations. A series of model tests for the deep-water lifting operation were performed at Ocean Engineering Basin of KRISO. For the model test, the vessel with a crane control system and a typical subsea manifold were examined. To validate the experimental results, a frequency-domain motion analysis method is applied. The coupled motion equations of the crane vessel and the lifted object are solved in the frequency domain with an additional linear stiffness matrix due to the hoisting wire. The hydrodynamic coefficients of the lifted object, which is a significant factor to affect the coupled dynamics, are estimated based on the perforation value of the structure and the CFD results. The discussions were made on three main points. First, the motion characteristics of the lifted object as well as the crane vessel were studied by comparing the calculation results. Second, the dynamic tension of the hoisting wire were evaluated under the various wave conditions. Final discussion was made on the effect of passive heave compensator on the motion and tension responses.

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Vacuum vessel for thermonuclear device

International Nuclear Information System (INIS)

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

1997-01-01

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

Vacuum vessel for thermonuclear device

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-31

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

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

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

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

Cold source vessel development for the advanced neutron source

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

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

International Nuclear Information System (INIS)

1999-01-01

on in-vessel debris coolability through inherent cooling mechanisms, FOREVER experiments on thermal and mechanical behaviour of a reactor pressure vessel during a severe accident, Experimental studies of heat transfer in the slotted channels at the CTF facility, Experimental study on CHF in a hemispherical narrow gap, Experiments on heat removal in a gap between debris crust and RPV wall), sub-session 4 (Creep behaviour of reactor pressure vessel lower head: Experimental investigation of creep behaviour of RPV lower head, Lower head thermo-mechanical behaviour, Pressure vessel creep rupture analysis, Parametric studies on creep behavior of a reactor pressure vessel lower head, Study of RPV materials with respect to mechanical behaviour in case of complete core fusion), sub-session 5 (Ex-vessel boiling and critical heat flux phenomena: Natural convection boiling on the outer surface of a hemispherical vessel surrounded by a thermal insulation structure, Reactor vessel external cooling for corium retention SULTAN experimental program and modelling with CATHARE code), and session 3 (Scaling to reactor severe accident conditions and reactor applications: Potential for in-vessel retention through ex-vessel flooding, In-vessel core melt retention by RPV external cooling for high power PWR MAAP4 analysis on a LBLOCA scenario without SI, Coupled thermal-hydraulic analyses of the molten pool and pressure vessel during a severe accident, Studies on core melt behaviour in a BWR pressure vessel lower head, Analysis of reactor lower head penetration tube failure, Thermal hydraulic and mechanical aspects of in-vessel retention of core debris)

Evaluation of buckling on containment metallic vessels

International Nuclear Information System (INIS)

Silveira, Renato Campos da; Mattar Neto, Miguel

2000-01-01

The buckling analysis represents one of the most important aspects of the structural projects of nuclear power plants containment metallic vessels and in this work the Case N-284-1 ASME Code is used for evaluation of those structures submitted to this failure mode. From the stress analysis, performed by using finite element method on discrete structures with shell elements, the procedure of the Code Case are applied to the evaluation of the containment metallic vessel of the Angra 2 nuclear power plant submitted to the own weight, seismic loads and uplift in case of accident. A study of pressure vessel reinforced by rings submit ed to the external pressure. Conclusions and commentaries are established based on the obtained results

Study air ingress into the reactor vessel using ICARE/CATHARE V2.0 in case of severe accident

International Nuclear Information System (INIS)

Gwenaelle Le Dantec; Fichot, F.

2005-01-01

Full text of publication follows: Safety analyses show that core degradation during a severe reactor accident would not be uniform. This was confirmed by TMI2 examinations. In fact, a central region of the core may overheat, melt and flow down to the lower plenum of the reactor while peripheral regions of the core would remain almost intact. Following rupture of the vessel by molten debris, air may be drawn from the containment by natural convection into the reactor coolant system, and react with the intact rods. Studying air ingress into the reactor vessel is of interest because the interaction of air with Zircaloy cladding can strongly affect the evolution of severe accident scenarios. The main effects are heat generation, increasing clad degradation, fission product release and nitriding. In case of air/steam recirculation in the vessel, significant nitriding of cladding can occur. The resulting ZrN phase is characterized by its brittleness and instability under oxidizing conditions, Oxidation of pre-existing ZrN phase layers has been observed to result in violent oxidation and heat release. Therefore, the first consequence for safety is a risk of strong deflagration in the vessel if a large number of rods on which a substantial layer of ZrN has grown are suddenly in contact with oxygen or steam. The second consequence is a late melting of core materials due to the very exothermic oxidation, leading to a late release of materials out of the reactor pressure vessel (RPV). In this paper we present an ICARE/CATHARE V2.0 calculation simulating air ingress into the vessel and in particular to describe the nitriding due to natural convection in the reactor vessel. The basic modeling and the necessary extensions of both ICARE and CATHARE are explained. The natural circulation is calculated to predict the regions of oxygen starvation where nitriding takes place. Key words: air ingress, nitriding, ICARE/CATHARE V2.0. (authors)

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.

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)

Experimental and Numerical Study of Effect of Thermal Management on Storage Capacity of the Adsorbed Natural Gas Vessel

KAUST Repository

Ybyraiymkul, Doskhan

2017-07-08

One of the main challenges in the adsorbed natural gas (ANG) storage system is the thermal effect of adsorption, which significantly lowers storage capacity. These challenges can be solved by efficient thermal management system. In this paper, influence of thermal management on storage capacity of the ANG vessel was studied experimentally and numerically. 3D numerical model was considered in order to understand heat transfer phenomena and analyze influence of thermal control comprehensively. In addition, a detailed 2D axisymmetric unit cell model of adsorbent layer with heat exchanger was developed, followed by optimization of heat exchanging device design to minimize volume occupied by fins and tubes. Heat transfer, mass transfer and adsorption kinetics, which occur in ANG vessel during charging process, are accounted for in models. Nelder-Mead method is implemented to obtain the geometrical parameters, which lead to the optimal characteristics of heat exchange. A new optimized configuration of ANG vessel was developed with compact heat exchanger. Results show that storage capacity of the ANG vessel increased significantly due to lowering of heat exchanger volume for 3 times from 13.5% to 4.3% and effective temperature control.

Atucha I nuclear power plant azimuthal ex-vessel flux profile evaluation

International Nuclear Information System (INIS)

Ferraro, Diego

2008-01-01

Irradiation damage in RPV (Reactor Pressure Vessel) in nuclear power plants is a key parameter to be analyzed in order to assess the plant integrity up to end of life and planning for a possible plant life extension. In this work a neutronic model in MCNP that represents a sector of 30 degrees of the Atucha I power plant nucleus has been consolidated with the results of an ex-vessel dosimetry made in the outer surface of the RPV s power plant in order to analyse the irradiation damage through the dpa rate. A strong dependents of the maximum point of damage with the loading of a peripheral channel was found, so a mitigation strategy was proposed, which is basically to empty this channel and its analogs in the rest of the nucleus. Analysing this second case a notable decrease of the damage is found in the zone considerated on the model (shown through the drop of de dpa rate in the zone). [es

Pediatric ureteropelvic junction obstruction: can magnetic resonance urography identify crossing vessels?

Energy Technology Data Exchange (ETDEWEB)

Parikh, Kushal R.; Kraft, Kate H.; Ivancic, Vesna; Smith, Ethan A.; Dillman, Jonathan R. [Section of Pediatric Radiology, Mott Children' s Hospital, Department of Radiology, University of Michigan Health System, Ann Arbor, MI (United States); Hammer, Matthew R. [University of Texas Southwestern, Department of Radiology, Dallas, TX (United States)

2015-11-15

MR Urography (MRU) is an increasingly used imaging modality for the evaluation of pediatric genitourinary obstruction. To determine whether pediatric MR urography (MRU) reliably detects crossing vessels in the setting of suspected ureteropelvic junction (UPJ) obstruction. The clinical significance of these vessels was also evaluated. We identified pediatric patients diagnosed with UPJ obstruction by MRU between May 2009 and June 2014. MRU studies were evaluated by two pediatric radiologists for the presence or absence of crossing vessels. Ancillary imaging findings such as laterality, parenchymal thinning/scarring, trapped fluid in the proximal ureter, and presence of renal parenchymal edema were also evaluated. Imaging findings were compared to surgical findings. We used the Mann-Whitney U test to compare continuous data and the Fisher exact test to compare proportions. Twenty-four of 25 (96%) UPJ obstructions identified by MRU were surgically confirmed. MRU identified crossing vessels in 10 of these cases, with 9 cases confirmed intraoperatively (κ = 0.92 [95% CI: 0.75, 1.0]). Crossing vessels were determined to be the primary cause of UPJ obstruction in 7/9 children intraoperatively, while in two children the vessels were deemed incidental and noncontributory to the urinary tract obstruction. There was no significant difference in age or the proportions of ancillary findings when comparing children without and with obstructing vessels. MRU allows detection of crossing vessels in pediatric UPJ obstruction. Although these vessels are the primary cause of obstruction in some children, they are incidental and non-contributory in others. Our study failed to convincingly identify any significant predictors (e.g., age or presence of renal parenchymal edema) that indicate when a crossing vessel is the primary cause of obstruction. (orig.)

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

International Nuclear Information System (INIS)

Muramatu, Toshiharu; Yamaguchi, Akira

2004-01-01

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

Nuclear reactor 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

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

A photoelastic study of the effects of an impulsive seismic wave on a nuclear containment vessel

International Nuclear Information System (INIS)

Burger, C.P.

1981-01-01

A dynamic photoelastic study of the progressive movement of a dilatational P-wave into a model of a nuclear containment vessel,is studied. The reflections at the dome abutments are observed and the strong flexural wave that deforms the dome itself is studied with photoelasticity and with dynamic strain gage procedures. (E.G.) [pt

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.)

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...

Planning of vessel grafts for reconstructive surgery in congenital heart diseases

Science.gov (United States)

Rietdorf, U.; Riesenkampff, E.; Schwarz, T.; Kuehne, T.; Meinzer, H.-P.; Wolf, I.

2010-02-01

The Fontan operation is a surgical treatment for patients with severe congenital heart diseases, where a biventricular correction of the heart can't be achieved. In these cases, a uni-ventricular system is established. During the last step of surgery a tunnel segment is placed to connect the inferior caval vein directly with the pulmonary artery, bypassing the right atrium and ventricle. Thus, the existing ventricle works for the body circulation, while the venous blood is passively directed to the pulmonary arteries. Fontan tunnels can be placed intra- and extracardially. The location, length and shape of the tunnel must be planned accurately. Furthermore, if the tunnel is placed extracardially, it must be positioned between other anatomical structures without constraining them. We developed a software system to support planning of the tunnel location, shape, and size, making pre-operative preparation of the tunnel material possible. The system allows for interactive placement and adjustment of the tunnel, affords a three-dimensional visualization of the virtual Fontan tunnel inside the thorax, and provides a quantification of the length, circumferences and diameters of the tunnel segments. The visualization and quantification can be used to plan and prepare the tunnel material for surgery in order to reduce the intra-operative time and to improve the fit of the tunnel patch.

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

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...

Fast Generation of Near-Optimal Plans for Eco-Efficient Stowage of Large Container Vessels

DEFF Research Database (Denmark)

Pacino, Dario; Delgado, Alberto; Jensen, Rune Møller

2011-01-01

Eco-efficient stowage plans that are both competitive and sustainable have become a priority for the shipping industry. Stowage planning is NP-hard and is a challenging optimization problem in practice. We propose a new 2-phase approach that generates near-optimal stowage plans and fulfills indus...

Stress analysis and evaluation of a rectangular pressure vessel

International Nuclear Information System (INIS)

Rezvani, M.A.; Ziada, H.H.; Shurrab, M.S.

1992-10-01

This study addresses structural analysis and evaluation of an abnormal rectangular pressure vessel, designed to house equipment for drilling and collecting samples from Hanford radioactive waste storage tanks. It had to be qualified according to ASME boiler and pressure vessel code, Section VIII; however, it had the cover plate bolted along the long face, a configuration not addressed by the code. Finite element method was used to calculate stresses resulting from internal pressure; these stresses were then used to evaluate and qualify the vessel. Fatigue is not a concern; thus, it can be built according to Section VIII, Division I instead of Division 2. Stress analysis was checked against the code. A stayed plate was added to stiffen the long side of the vessel

Crack growth rates in vessel head penetration materials

International Nuclear Information System (INIS)

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

1994-01-01

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

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)

Vessel abnormalization: another hallmark of cancer? Molecular mechanisms and therapeutic implications.

Science.gov (United States)

De Bock, Katrien; Cauwenberghs, Sandra; Carmeliet, Peter

2011-02-01

As a result of excessive production of angiogenic molecules, tumor vessels become abnormal in structure and function. By impairing oxygen delivery, abnormal vessels fuel a vicious cycle of non-productive angiogenesis, which creates a hostile microenvironment from where tumor cells escape through leaky vessels and which renders tumors less responsive to chemoradiation. While anti-angiogenic strategies focused on inhibiting new vessel growth and destroying pre-existing vessels, clinical studies showed modest anti-tumor effects. For many solid tumors, anti-VEGF treatment offers greater clinical benefit when combined with chemotherapy. This is partly due to a normalization of the tumor vasculature, which improves cytotoxic drug delivery and efficacy and offers unprecedented opportunities for anti-cancer treatment. Here, we overview key novel molecular players that induce vessel normalization. Copyright © 2010 Elsevier Ltd. All rights reserved.

45 CFR 673.5 - Emergency response plan.

Science.gov (United States)

2010-10-01

... ensure that: (a) The vessel owner's or operator's shipboard oil pollution emergency plan, prepared and... Pollution from Ships, 1973, as modified by the Protocol of 1978 relating thereto (MARPOL 73/78), has provisions for prompt and effective response action to such emergencies as might arise in the performance of...

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.

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.)

U.S. and French approaches to reactor pressure vessel integrity

International Nuclear Information System (INIS)

Griesbach, T.J.; Buchalet, C.; Server, W.L.

1990-01-01

The effects of radiation embrittlement on the reactor pressure vessel must be considered for continued safe operation of nuclear power plants. The consequences of radiation embrittlement require detailed assessments of the margins of safety against brittle fracture of the vessel. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code and U.S. Regulations often use conservative approaches for these assessments which can eventually lead to severe operational hardships for some plants. Taking a look at alternative integrity approaches, such as those demonstrated in France, could ultimately result in improved ASME Code and Regulatory limits. The French studies have shown the significance of performing proper in- service inspections to reliably show that no defects larger than a predetermined size (or class) exist in the inspected region of a vessel. The predetermined size is based upon previous studies on the types of manufacturing defects which can potentially exist in French vessels. Enhanced linear elastic and elastic-plastic fracture mechanics methodologies can be applied to evaluate such defects to assure that brittle fracture will not occur

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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.

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...

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.)

Westinghouse experience in using mechanical cutting for reactor vessel internals segmentation

International Nuclear Information System (INIS)

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

2010-01-01

Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques. Mechanical cutting has been used by Westinghouse since 1999 for both PWRs and BWRs and its process has been continuously improved over the years. Detailed planning is essential to a successful project, and typically a 'Segmentation and Packaging Plan' is prepared to document the effort. The usual method is to start at the end of the process, by evaluating the waste disposal requirements imposed by the waste disposal agency, what type and size of containers are available for the different disposal options, and working backwards to select the best cutting tools and finally the cut geometry required. These plans are made utilizing advanced 3-D CAD software to model the process. Another area where the modelling has proven invaluable is in determining the logistics of component placement and movement in the reactor cavity, which is typically very congested when all the internals are out of the reactor vessel in various stages of segmentation. The main objective of the segmentation and packaging plan is to determine the strategy for separating the highly activated components from the less activated material, so that they can be disposed of in the most cost effective manner. Usually, highly activated components cannot be shipped off-site, so they must be packaged such that they can be dry stored with the spent fuel in an Independent Spent Fuel Storage Installation (ISFSI). Less activated components can be shipped to an off-site disposal site depending on space availability. Several of the

Appraisal of Emissions from Ocean-going Vessels Coming to Kolkata Port, India

Science.gov (United States)

Mandal, Anindita; Biswas, Jhumoor; Roychowdhury, Soma; Farooqui, Zuber M.

2017-12-01

Rapid economic growth has escalated India's share in international trade. The pressure on these ports, which handle a substantial portion of the trade, has increased to perform with optimal efficiency, and decrease turnaround time so as to increase the number of ships visiting the port area. The caveat is that increased shipping activity is accompanied by enhanced emissions of harmful pollutants and green house gases. This study has revealed increased turnaround time for ships resulting in substantial emissions from auxiliary engines. There should be an optimum balance between operational control and environmental control of pollutants. Kolkata is a megacity with active riverine ports that can generate high levels of air quality emissions, especially NOx, SOx and particulate matter. An exhaustive annual emissions inventory based on ocean going vessels activity has been developed for 2013-2014 for Kolkata port, using recent EPA approved methodology. This includes greenhouse gas emissions from marine engines as well. The study indicates that amongst the different categories of ocean going ships, containers contribute the most (49%) of air and greenhouse gas emissions in 75th percentile class and above followed by general cargo (14%) and oil tankers (13%). The study depicts existing status of marine emissions in Kolkata port from ocean going vessels, which would serve in development of integrated air quality and climate change management plans and serve as a prototype for other major ports of India.

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...

THE RELATIONSHIP BETWEEN FOVEAL AVASCULAR ZONE AREA, VESSEL DENSITY, AND CYSTOID CHANGES IN DIABETIC RETINOPATHY, AN OPTICAL COHERENCE TOMOGRAPHY ANGIOGRAPHY STUDY.

Science.gov (United States)

Tarassoly, Kia; Miraftabi, Arezoo; Soltan Sanjari, Mostafa; Parvaresh, Mohammad Mehdi

2017-06-29

To measure the foveal avascular zone (FAZ) areas and vessel densities of patients with diabetic retinopathy and to study their relationship with diabetic cystoid changes and retinal thickness. Prospective case series of 51 eyes of 31 patients with diabetic retinopathy. The eyes were grouped based on the presence or absence of cystoid edema and evaluated using optical coherence tomography angiography. The FAZ areas and vessel density were compared. The FAZ area at the superficial capillary plexus level was equal between the eyes with and without cystoid edema. Vessel density did not differ as well. There was no correlation with retinal thickness. In eyes with cystoid changes, FAZ area changes at the deep capillary plexus level were difficult to interpret. The FAZ area and vessel density at the superficial capillary plexus level are reproducible and independent of the presence of cystoid edema.

Standard High Solids Vessel Design De-inventory Simulant Qualification

Energy Technology Data Exchange (ETDEWEB)

Fiskum, Sandra K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burns, Carolyn A.M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Linn, Diana T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Peterson, Reid A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smoot, Margaret R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-09-12

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is working to develop a Standard High Solids Vessel Design (SHSVD) process vessel. To support testing of this new design, WTP engineering staff requested that a Newtonian simulant be developed that would represent the de-inventory (residual high-density tank solids cleanout) process. Its basis and target characteristics are defined in 24590-WTP-ES-ENG-16-021 and implemented through PNNL Test Plan TP-WTPSP-132 Rev. 1.0. This document describes the de-inventory Newtonian carrier fluid (DNCF) simulant composition that will satisfy the basis requirement to mimic the density (1.18 g/mL ± 0.1 g/mL) and viscosity (2.8 cP ± 0.5 cP) of 5 M NaOH at 25 °C.1 The simulant viscosity changes significantly with temperature. Therefore, various solution compositions may be required, dependent on the test stand process temperature range, to meet these requirements. Table ES.1 provides DNCF compositions at selected temperatures that will meet the density and viscosity specifications as well as the temperature range at which the solution will meet the acceptable viscosity tolerance.

Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

International Nuclear Information System (INIS)

Bros, J.; Ballesteros, A.; Lopez, A.

1993-01-01

Commercial pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear power plants contain a series of pressure vessel steel surveillance capsules as the principal means of monitoring radiation effects on the pressure vessel. Changes in fracture toughness are more severe in surveillance capsules than in reactor vessel materials because of their proximity of the reactor core. Therefore, it is possible to predict changes in fracture toughness of the reactor vessel materials. This paper describes the status of the reactor vessel surveillance program relating to Spanish nuclear power plants. To date, twelve capsules have been removed and analyzed from seven of the nine Spanish reactors in operation. The results obtained from the analysis of these capsules are compared with the predictions of the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, by means of measured and expected increase of the nil-ductility transition reference temperature (RT NDT ). The comparison is made considering the different variables normally included in the studies of radiation response of reactor pressure vessel materials, such as copper content of steel, level of neutron fluence above 1 MeV, base metal or weld metal, and so forth. The surveillance data have been used for determining the adjusted reference temperatures and upper shelf energies at any time. The results have shown that the seven pressure vessels are in excellent condition to continue operating with safety against brittle fracture beyond the design life, without the need to recuperate the degraded properties of the materials by annealing of the vessel

Diaphragmatic lymphatic vessel behavior during local skeletal muscle contraction.

Science.gov (United States)

Moriondo, Andrea; Solari, Eleonora; Marcozzi, Cristiana; Negrini, Daniela

2015-02-01

The mechanism through which the stresses developed in the diaphragmatic tissue during skeletal muscle contraction sustain local lymphatic function was studied in 10 deeply anesthetized, tracheotomized adult Wistar rats whose diaphragm was exposed after thoracotomy. To evaluate the direct effect of skeletal muscle contraction on the hydraulic intraluminal lymphatic pressures (Plymph) and lymphatic vessel geometry, the maximal contraction of diaphragmatic fibers adjacent to a lymphatic vessel was elicited by injection of 9.2 nl of 1 M KCl solution among diaphragmatic fibers while Plymph was recorded through micropuncture and vessel geometry via stereomicroscopy video recording. In lymphatics oriented perpendicularly to the longitudinal axis of muscle fibers and located at skeletal muscle contraction (Dmc) decreased to 61.3 ± 1.4% of the precontraction value [resting diameter (Drest)]; however, if injection was at >900 μm from the vessel, Dmc enlarged to 131.1 ± 2.3% of Drest. In vessels parallel to muscle fibers, Dmc increased to 122.8 ± 2.9% of Drest. During contraction, Plymph decreased as much as 22.5 ± 2.6 cmH2O in all submesothelial superficial vessels, whereas it increased by 10.7 ± 5.1 cmH2O in deeper vessels running perpendicular to contracting muscle fibers. Hence, the three-dimensional arrangement of the diaphragmatic lymphatic network seems to be finalized to efficiently exploit the stresses exerted by muscle fibers during the contracting inspiratory phase to promote lymph formation in superficial submesothelial lymphatics and its further propulsion in deeper intramuscular vessels. Copyright © 2015 the American Physiological Society.

Absorbed dose calculations to blood and blood vessels for internally deposited radionuclides

International Nuclear Information System (INIS)

Akabani, G.; Poston, J.W. Sr.

1992-01-01

At present, absorbed dose calculations for radionuclides in the human circulatory system use relatively simple models and are restricted in their applications. To determine absorbed doses to the blood and to the surface of the blood vessel wall, Monte Carlo calculations were performed using the code Electron Gamma Shower (EGS4). Absorbed doses were calculated for the blood and the blood vessel wall (lumen) for different blood vessel sizes. The radionuclides chosen for this study were those commonly used in nuclear medicine. No diffusion of the radionuclide into the blood vessel was or cross fire between blood vessels was assumed. Results are useful in assessing the doses to blood and blood vessel walls for different nuclear medicine procedures

Absorbed dose calculations to blood and blood vessels for internally deposited radionuclides

International Nuclear Information System (INIS)

Akabani, G.; Poston, J.W.

1991-05-01

At present, absorbed dose calculations for radionuclides in the human circulatory system used relatively simple models and are restricted in their applications. To determine absorbed doses to the blood and to the surface of the blood vessel wall, EGS4 Monte Carlo calculations were performed. Absorbed doses were calculated for the blood and the blood vessel wall (lumen) for different blood vessels sizes. The radionuclides chosen for this study were those commonly used in nuclear medicine. No diffusion of the radionuclide into the blood vessel was assumed nor cross fire between vessel was assumed. Results are useful in assessing the dose in blood and blood vessel walls for different nuclear medicine procedures. 6 refs., 6 figs., 5 tabs

A study of the external cooling capability for the prevention of reactor vessel failure

Energy Technology Data Exchange (ETDEWEB)

Chang, S H; Baek, W P; Moon, S K; Yang, S H; Kim, S H [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1995-07-15

This study (a 3-year program) aims to perform a comprehensive assessment of the feasibility of external vessel flooding with respect to advanced pressurized water reactor plants to be built in Korea. During the second year, appropriate correlations have been chosen to describe the phenomena resulted from the external flooding on the basis of review works. Also performed is to develop the computer program using the chosen correlations and to accomplish the thermal analysis for assessment of the cooling capability of external flooding. Accomplished works for second year are as follows. Review of analytical and experimental works related to the external flooding are performed, appropriate correlations are chosen to describe the phenomena resulted from the external flooding on the basis of first and second year review works. A computer program is also developed to predict the temperature distribution of reactor vessel lower head. Thermal analyses are performed to judge the feasibility of external flooding using developed computer program.

Design study of a new vacuum vessel for Doublet III

International Nuclear Information System (INIS)

Rawls, J.M.; Davis, L.G.; Anderson, P.M.

1980-10-01

The principal thrust of the project was to examine a single design in enough depth to gain confidence in the feasibility and desirability of specific design features. However, a valuable spin-off of the project was to develop information of a more generic character to aid in future studies of possibilities for Doublet III. For example, we now feel that Doublet III can be reconfigured with any of a variety of new vacuum vessels, poloidal coil sets, and auxiliary heating systems within three years of project initiation, a period that is short compared to the time scale for developing a completely new facility. In addition, this can be accomplished at a fraction of the cost required to develop a comparable facility

Automatic blood vessel based-liver segmentation using the portal phase abdominal CT

Science.gov (United States)

Maklad, Ahmed S.; Matsuhiro, Mikio; Suzuki, Hidenobu; Kawata, Yoshiki; Niki, Noboru; Shimada, Mitsuo; Iinuma, Gen

2018-02-01

Liver segmentation is the basis for computer-based planning of hepatic surgical interventions. In diagnosis and analysis of hepatic diseases and surgery planning, automatic segmentation of liver has high importance. Blood vessel (BV) has showed high performance at liver segmentation. In our previous work, we developed a semi-automatic method that segments the liver through the portal phase abdominal CT images in two stages. First stage was interactive segmentation of abdominal blood vessels (ABVs) and subsequent classification into hepatic (HBVs) and non-hepatic (non-HBVs). This stage had 5 interactions that include selective threshold for bone segmentation, selecting two seed points for kidneys segmentation, selection of inferior vena cava (IVC) entrance for starting ABVs segmentation, identification of the portal vein (PV) entrance to the liver and the IVC-exit for classifying HBVs from other ABVs (non-HBVs). Second stage is automatic segmentation of the liver based on segmented ABVs as described in [4]. For full automation of our method we developed a method [5] that segments ABVs automatically tackling the first three interactions. In this paper, we propose full automation of classifying ABVs into HBVs and non- HBVs and consequently full automation of liver segmentation that we proposed in [4]. Results illustrate that the method is effective at segmentation of the liver through the portal abdominal CT images.

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.

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)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Two, three and four buttressed PWR containment vessels. A comparative study

International Nuclear Information System (INIS)

Dufour, C.J.; Cheyrezy, M.H.; Thorsen, N.E.

1977-01-01

The purpose of the paper is to analyse the advantages and drawbacks of different arrangements of hood tendons and buttresses for a PWR containment vessel. It is shown that the solution with two buttresses and full hoop tendons (through 360 0 ) gives: acceptable secondary stresses and strains; a lower total cost of the prestressing; a shorter time schedule for prestressing operations; a smaller 'forbidden area' for the penetration lay out. This arrangement has been used for the first time (to our knowledge) in PWR containment design for the nuclear plant of DOEL III, Belgium. The comparative study has been carried out for a containment vessel with a 85 cm thick wall and an interior diameter of 42.50 m. The Guaranteed Ultimate Tensile Strength of the tendons is 9000 KN. The minimum required prestressing force is 10 300 KN per linear meter of height. The buttresses are 4.0 m long and 1.70 m thick on their vertical centerline. The following arrangements have been studied: 4 buttresses and 3/4 hoop tendons; 3 buttresses and 2/3 hoop tendons; 3 buttresses and full hoop tendons; 2 buttresses and full hoop tendons. The ovalisation and other secondary effects result mainly from three different causes: the average prestressing force is not constant around a hoop; the buttress itself constitutes a sudden thickening and creates local disturbances even under an axisymmetric loading: the anchored tendons are straight over their first 6 meters and they initiate local stress perturbations in the vicinity of the junction of the buttress and the wall. The determination of these secondary effects has been performed by a plane stress finite element analysis

Pericyte coverage of abnormal blood vessels in myelofibrotic bone marrows

DEFF Research Database (Denmark)

Zetterberg, Eva; Vannucchi, Alessandro M; Migliaccio, Anna Rita

2007-01-01

BACKGROUND AND OBJECTIVES: Myelofibrotic bone marrow displays abnormal angiogenesis but the pathogenic mechanisms of this are poorly understood. Since pericyte abnormalities are described on solid tumor vessels we studied whether vessel morphology and pericyte coverage in bone marrow samples from...

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)

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

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

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

Design criteria and pressure vessel codes - an American view

International Nuclear Information System (INIS)

Tuppeny, W.H.

1975-01-01

To the pressure vessel designer, codes and criteria represent the common ground where the stress analyst and the metallurgist must interact and evolve rules and procedures which will ensure safety and open-ended responsiveness to technological, economic, and environmental change. The paper briefly discusses the evolution and rationale behind the current ASME code sections -emphasizing those portions applicable to designs operating in the creep range. The author then proposes a plan of action so that the analysts and materials people can make optimum use of time and resources, and evolve data and design criteria which will be responsive to changing technology and the economic and safety requirements of the future. (author)

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

CSIR Research Space (South Africa)

Tulsi, K

2014-05-01

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

Recovery of testicular blood flow following ligation of testicular vessels

International Nuclear Information System (INIS)

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

1989-01-01

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

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

Embrittlement and annealing of reactor pressure vessel steels: comparison of BR3 surveillance and vessel plates to the surrogate plates representative of the Yankee Rowe vessel

International Nuclear Information System (INIS)

Fabry, A.; Chaouadi, T.; Puzzolante, J.L.; Van de Velde, J.; Biemiller, E.C.; Rossinski, S.T.; Carter, R.G.

1996-07-01

The sister pressure vessels at the BR3 and Rowe Yankee PWR plants were operated at a lower-than-usual temperature (260 degrees Celsius) and their plates were austenitized at higher-than-usual temperature (970 degrees Celsius). A heat tratemement leading to a coarser microstructure than typical for the fine grain plates that are considered in development of USNRC Regulatory guide 1.99. This material displayed outlier behaviour charackterized by a 41J CVN-shift significantly larger than predicted by Regulatory Guide 1.99. Because lower radiation temperature and nickell alloying are generally considered detrimental to irradiation sensitivity, there was a major concern that the nickel-modified lower Rowe plate and the nickel-modified BR3 plate may become too embrittled to satisfy the toughness requirements enbodied in the PTS screening criterion. This paper compares three complementary studies undertaken to clarify these uncertainties: 1) the accelerated irradiation and test program launched in 1990 by Yankee Atomic Electric Company using typical vessel plate materials containing 0.24% copper at two nickel levels: YA1, 0.63% (A533-B) and YA9, 0.19% (A302-B). These were heat-treated to produce the coarse and fine grain microstructures representative of the Yankee/BR3 and the Regulatory Guide plates, respectively, 2) the BR3 surveillance and vessel testing program: this vessel was wet-annealed in 1984, relicensed for operation till the plant shutdown in 1987, ANCL was trepanned in early 1995, 3) the accelerated irradiations in the Belgian BR2 test reactor of the Yankee coarse grain plates YA1 and YA9 together with BR3 vessel specimens extracted at nozzle elevation, a location with negligible radiation exposure. It is shown that the PTS screening criterion was never attained by the BR3 and Rowe plates, and that the BR3 vessel anneal was neither necessary nor sufficient. Finally, the sensitivity of embrittlement, annealing and post-annealing reembrittlement to irradiation

Embrittlement and annealing of reactor pressure vessel steels: comparison of BR3 surveillance and vessel plates to the surrogate plates representative of the Yankee Rowe vessel

Energy Technology Data Exchange (ETDEWEB)

Fabry, A.; Chaouadi, T.; Puzzolante, J.L.; Van de Velde, J. [Centre de l``Etude de l``Energie Nucleaire, Mol (Belgium); Biemiller, E.C. [Yankee Atomic Electric Company, Bolton (United States); Rossinski, S.T.; Carter, R.G. [Electric Power Research Institute, Charlotte (United States)

1996-07-01

The sister pressure vessels at the BR3 and Rowe Yankee PWR plants were operated at a lower-than-usual temperature (260 degrees Celsius) and their plates were austenitized at higher-than-usual temperature (970 degrees Celsius). A heat tratemement leading to a coarser microstructure than typical for the fine grain plates that are considered in development of USNRC Regulatory guide 1.99. This material displayed outlier behaviour charackterized by a 41J CVN-shift significantly larger than predicted by Regulatory Guide 1.99. Because lower radiation temperature and nickell alloying are generally considered detrimental to irradiation sensitivity, there was a major concern that the nickel-modified lower Rowe plate and the nickel-modified BR3 plate may become too embrittled to satisfy the toughness requirements enbodied in the PTS screening criterion. This paper compares three complementary studies undertaken to clarify these uncertainties: 1) the accelerated irradiation and test program launched in 1990 by Yankee Atomic Electric Company using typical vessel plate materials containing 0.24% copper at two nickel levels: YA1, 0.63% (A533-B) and YA9, 0.19% (A302-B). These were heat-treated to produce the coarse and fine grain microstructures representative of the Yankee/BR3 and the Regulatory Guide plates, respectively, 2) the BR3 surveillance and vessel testing program: this vessel was wet-annealed in 1984, relicensed for operation till the plant shutdown in 1987, ANCL was trepanned in early 1995, 3) the accelerated irradiations in the Belgian BR2 test reactor of the Yankee coarse grain plates YA1 and YA9 together with BR3 vessel specimens extracted at nozzle elevation, a location with negligible radiation exposure. It is shown that the PTS screening criterion was never attained by the BR3 and Rowe plates, and that the BR3 vessel anneal was neither necessary nor sufficient. Finally, the sensitivity of embrittlement, annealing and post-annealing reembrittlement to irradiation

Production management and quality assurance for the fabrication of the In-Vessel Components of the stellarator Wendelstein 7-X

Energy Technology Data Exchange (ETDEWEB)

Li, C., E-mail: chuanfei.li@ipp.mpg.de; Boscary, J.; Dekorsy, N.; Junghanns, P.; Mendelevitch, B.; Peacock, A.; Pirsch, H.; Sellmeier, O.; Springer, J.; Stadler, R.; Streibl, B.

2014-10-15

Highlights: • Thousand parts for the divertor, first wall, cooling supply and diagnostics as W7-X In-Vessel Components. • Database building including part and assembly data, work and capacity organization, quality assurance documents. • Production management system to organize the fabrication and the associated quality assurance. • Successful use of an efficient and flexible product planning and scheduling tool for W7-X In-Vessel Components. - Abstract: The In-Vessel Components (IVC) of the stellarator Wendelstein 7-X consist of the divertor components and the first wall (FW) with their internal water cooling supply and a set of diagnostics. Due to the significant amount of different components, including many variants, a tool called Production Managing System (PMS) has been developed to organize the fabrication and the associated quality assurance. The PMS works by building a database containing the basic parts and assembly data, manufacturing and quality control plans, and available machine capacity. The creation of this database is based mainly on the parts lists, the manufacturing drawings, and details of the working flow organization. As a consequence of the learning process and technical adjustments during the design and manufacturing phase, the database needed to be permanently updated. Therefore an interface tool to optimize the data preparation has been developed. PMS has been demonstrated to be an efficient tool to support the IVC production activities providing reliable planning estimates, easily adaptable to problems encountered during the fabrication and provided a basis for the integration of quality assurance requirements.

Production management and quality assurance for the fabrication of the In-Vessel Components of the stellarator Wendelstein 7-X

International Nuclear Information System (INIS)

Li, C.; Boscary, J.; Dekorsy, N.; Junghanns, P.; Mendelevitch, B.; Peacock, A.; Pirsch, H.; Sellmeier, O.; Springer, J.; Stadler, R.; Streibl, B.

2014-01-01

Highlights: • Thousand parts for the divertor, first wall, cooling supply and diagnostics as W7-X In-Vessel Components. • Database building including part and assembly data, work and capacity organization, quality assurance documents. • Production management system to organize the fabrication and the associated quality assurance. • Successful use of an efficient and flexible product planning and scheduling tool for W7-X In-Vessel Components. - Abstract: The In-Vessel Components (IVC) of the stellarator Wendelstein 7-X consist of the divertor components and the first wall (FW) with their internal water cooling supply and a set of diagnostics. Due to the significant amount of different components, including many variants, a tool called Production Managing System (PMS) has been developed to organize the fabrication and the associated quality assurance. The PMS works by building a database containing the basic parts and assembly data, manufacturing and quality control plans, and available machine capacity. The creation of this database is based mainly on the parts lists, the manufacturing drawings, and details of the working flow organization. As a consequence of the learning process and technical adjustments during the design and manufacturing phase, the database needed to be permanently updated. Therefore an interface tool to optimize the data preparation has been developed. PMS has been demonstrated to be an efficient tool to support the IVC production activities providing reliable planning estimates, easily adaptable to problems encountered during the fabrication and provided a basis for the integration of quality assurance requirements

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.

Eddy current testing of composite pressure vessels

Science.gov (United States)

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

2018-04-01

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

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

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

Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 3. Numerical investigation for thermal stratification phenomena in the upper plenum

International Nuclear Information System (INIS)

Muramatsu, Toshiharu; Yamaguchi, Akira

2002-06-01

A large-scale sodium-cooled fast breeder reactor in the feasibility studies on commercialized fast reactors has a feature of consideration of thorough simplified and compacted systems and components design to realize drastic economical improvements. Therefore, special attentions should be paid to thermohydraulic designs for gas entrainment behavior from free surface, flow-induced vibration of in-vessel components, thermal stratification in the plenum, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermal stratification characteristics in the upper plenum, and to investigate trade-off relations between gas entrainment and thermal stratification phenomena on in-vessel structures for the elimination of gas entrainment possibility. From the analysis, the following results were obtained. (1) Dummy plug insertion to a slit of the upper core structure is one of the effective measures to stabilize the in-vessel flow patterns and to mitigate in-vessel thermal shocks. (2) Though flow guide device such as a baffle ring attached to reactor vessel wall is an effective measure to eliminate impinging jet to dipped plate, rising characteristics of the thermal stratification interface are affected by the baffle ring devise. (3) Thermal stratification characteristics are not influenced very much by the installation of a partial inner barrel to the dipped plate, which is an effective measure to reduce the horizontal flow velocity components at free surface. (4) Labyrinth structures to the gap between the reactor vessel wall and the outer dipped plate have direct effects upon in-vessel thermal shock characteristics including thermal stratification phenomena due to the closing of flow path between the upper plenum and the free surface plenum. (author)

Application of annealing for extension of WWER vessel lives

International Nuclear Information System (INIS)

Badanin, V.; Dragunow, Yu.G.; Fedorov, V.; Gorynin, I.; Nickolaev, V.

1992-01-01

The safe operation of nuclear power plants (NPP) is dependent upon the assurance that the reactor pressure vessel will not fail in a brittle manner when the effects of radiation embrittlement are taken into account. The recovery of the properties of the irradiated materials is an important way of extending the operating life of a reactor vessel. The intent of this paper is to demonstrate the efficiency of thermal annealing for the recovery of reactor vessel material properties and to present the implications for extended service life. In order to substantiate the application of annealing to the extensior of the service life of vessels, detailed investigations were conducted which involved thermal annealing temperature and time, fast neutron fluence, and metallurgical factors (i.e. impurity contents) on the recovery of properties after the annealing of irradiated materials. Similar studies were continued to determine predictive methods for radiation embrittlement after repeated annealings. In May 1987 the first pilot annealing of a commercial reactor vessel (Novo-Voronezhskaya, III, NPP) was performed. The development of the annealing equipment and investigations performed to test the annealing process proved successful, and an improved safe operation for the reactor vessel was thus atttained providing for an extended service life. (orig.)

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

75 FR 12250 - Notification of the Imposition of Conditions of Entry for Certain Vessels Arriving to the United...

Science.gov (United States)

2010-03-15

..., 2010. ADDRESSES: This notice will be available for inspection and copying at the Docket Management... conducted pursuant to the provisions of 46 U.S.C. 70108 and the International Ship and Port Facility... five port calls. Vessels must: Implement measures per the ship's security plan equivalent to ``Security...

Mark III Containment vessel/annulus concrete design

International Nuclear Information System (INIS)

Chang, P.S.; Moussa, M.M.

1981-01-01

Recently, engineers have been considering the significant dynamic impact of safety/relief valve (S/RV) discharge loads on the containment structures, safety equipment, and piping systems in BWR type reactors. For a plant in the construction stage, extensive modifications will be made to qualify these new loads. The lower portion of the containment vessel serves as a suppression pool pressure boundary and is designed to sustain the effects of postulated loss of coolant accidents, seismic occurrences, S/RV discharge loads, and other effects. Extremely high spectral peak accelerations of the free-standing steel containment vessel can be obtained during the air dearing process of the S/RV discharge. Parametric studies indicated that a substantial reduction in response can be obtained by increasing the stiffness of the steel containment vessel in the lover area. A concrete backing configuration in the suppression pool area of Mark III Containment is proposed in this paper. A composite action is assumed between the steel containment vessel shell and the concrete section. The system is physically separated from the shield building. This approach warrants an early erection of the shield building and a late installation of piping systems in the containment vessel suppression pool area. Finite element analyses are performed by using ASHSD2 and EASE2 computer codes. The results of the analyses have shown the proposed stress criteria are satisfied. The approach pressented is justified to be a workable system for a new plant design. (orig./HP)

Benchmark study of shear buckling of a cylindrical vessel. Part 2

International Nuclear Information System (INIS)

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

1988-01-01

In Liquid Metal Fast Breeder Reactors (LMFBR) potential shear buckling failures of the primary vessel, induced through seismic excitations, have to be considered. The primary vessel material, typically 316 stainless steel, has a low yield strength at the normal operating temperatures of around 400 0 C to 500 0 C. There characteristics tend to make the structure relatively flexible and subject to potential elasto-plastic shear buckling failure. The use of finite element techniques in buckling analyses is currently becoming more accepted. There are at present many finite element codes available which have the capacibility to solve buckling problems. The objective of the study reported herein was to follow on from the previous code validation exercise and investigate the ability of finite element codes to predict buckling behaviour in another test cylinder [a/h = 83, a/L = 1] where non-linear effects would be more significant and plastic shear buckling could be a failure mode. As before four organisations took part in the code validation exercise. NNC [UK] and ISMES [Italy] used the commercially available general purpose FE code ABAQUS. CEA [France] used INCA and BILBO which are members of the commercially available CASTEM suite of FE program. Novatome [France] used their in-house FE code NOVNL. The joint effort was co-ordinated by NNC with the assistance of the Commission of the European Communities Working on Codes and Standards AG2

Fluid and mass transport in a single lymphatic blood vessel

International Nuclear Information System (INIS)

Bestman, A.R.

1987-08-01

The problem considers the single blood vessel model in pulmonary circulation in the presence of gravitation and mass transfer. The tissue surrounding the blood vessel is modelled as a permeable medium distinct from the blood vessel which is a normal free space. On the assumption that the mass concentration varies slowly at the interface between the blood vessel and the tissue, the problem is tackled by asymptotic approximation. A crucial point of the analysis is the dependence of the flow variables on the permeability K of the tissue in a completely arbitrary manner. A primary conjecture of the study is the intimacy of the pathological pulmonary edema and the parameter K. (author). 4 refs

Efficient Vessel Tracking  with Accuracy Guarantees

DEFF Research Database (Denmark)

Redoutey, Martin; Scotti, Eric; Jensen, Christian Søndergaard

2008-01-01

Safety and security are top concerns in maritime navigation, particularly as maritime traffic continues to grow and as crew sizes are reduced. The Automatic Identification System (AIS) plays a key role in regard to these concerns. This system, whose objective is in part to identify and locate ves...... accuracies at lower communication costs. The techniques employ movement predictions that are shared between vessels and the VTS. Empirical studies with a prototype implementation and real vessel data demonstrate that the techniques are capable of significantly improving the AIS....

Internal Friction of Pressure Vessel Steel Embrittlement

International Nuclear Information System (INIS)

Van Ouytsel, K.

2001-01-01

The contribution consists of an abstract of a PhD thesis. The thesis contains a literature study, a description of the construction details of a new inverted torsion pendulum. This device was designed to investigate pressure-vessel steels at high amplitudes (10 -4 to 10 -2 ) and over a wide temperature range (90-700K) at approximately 1 Hz in the irradiated condition. Results of measurements on a variety of reactor pressure vessel steels by means of the torsion penduli are reported and interpreted

Noise exposure and cognitive performance: A study on personnel on board Royal Norwegian Navy vessels.