Hierarchical and coupling model of factors influencing vessel traffic flow.

Science.gov (United States)

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

2017-01-01

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

AIS data based vessel speed, course and path analysis in the Botlek area in the Port of Rotterdam

NARCIS (Netherlands)

Shu, Y.; Daamen, W.; Ligteringen, H.; Hoogendoorn, S.P.

2012-01-01

Maritime traffic safety and port capacity is increasingly important nowadays. Due to the fast development of vessel traffic in ports and waterways, a lot of attention has been paid to maritime traffic safety and port capacity. Many simulation models have been used to predict traffic safety and port

Hierarchical and coupling model of factors influencing vessel traffic flow.

Directory of Open Access Journals (Sweden)

Zhao Liu

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

Development of FB-MultiPier dynamic vessel-collision analysis models, phase 2.

Science.gov (United States)

2014-07-01

Massive waterway vessels such as barges regularly transit navigable waterways in the U.S. During passages that fall within : the vicinity of bridge structures, vessels may (under extreme circumstances) deviate from the intended vessel transit path. A...

33 CFR 165.114 - Safety and Security Zones: Escorted Vessels-Boston Harbor, Massachusetts.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety and Security Zones... COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PORTS AND WATERWAYS SAFETY REGULATED NAVIGATION... Guard District § 165.114 Safety and Security Zones: Escorted Vessels—Boston Harbor, Massachusetts. (a...

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

Conditions of using floating cranes for lifting sunken objects on inland waterways

OpenAIRE

Slobodan M. Radojević

2012-01-01

This paper presents the conditions for using floating cranes for lifting sunken vessels and other objects on inland waterways. Basic technical data are given together with technical details for the usage of floa ting cranes for lifting sunken objects. The paper points to the importance of lifting sunken objects and their removal from inland waterways in the Republic of Serbia.

Conditions of using floating cranes for lifting sunken objects on inland waterways

Directory of Open Access Journals (Sweden)

Slobodan M. Radojević

2012-04-01

Full Text Available This paper presents the conditions for using floating cranes for lifting sunken vessels and other objects on inland waterways. Basic technical data are given together with technical details for the usage of floa ting cranes for lifting sunken objects. The paper points to the importance of lifting sunken objects and their removal from inland waterways in the Republic of Serbia.

76 FR 45693 - Safety Zone; San Diego POPS Fireworks, San Diego, CA

Science.gov (United States)

2011-08-01

...-AA00 Safety Zone; San Diego POPS Fireworks, San Diego, CA AGENCY: Coast Guard, DHS. ACTION: Temporary... San Diego Bay in support of the San Diego POPS Fireworks. This safety zone is necessary to provide for... of the waterway during scheduled fireworks events. Persons and vessels will be prohibited from...

76 FR 36318 - Safety Zone; Waterway Closure, Morgan City-Port Allen Route From Mile Marker 0 to Port Allen Lock

Science.gov (United States)

2011-06-22

... Executive Order 13045, Protection of Children from Environmental Health Risks and Safety Risks. This rule is not an economically significant rule and does not create an environmental risk to health or risk to...-AA00 Safety Zone; Waterway Closure, Morgan City-Port Allen Route From Mile Marker 0 to Port Allen Lock...

78 FR 29094 - Safety Zone; Coronado Fourth of July Fireworks, Glorietta Bay; Coronado, CA

Science.gov (United States)

2013-05-17

... waterways, darkness punctuated by bright flashes of light and burning debris has the potential to result in...: The safety zone will only be in effect for one hour, late in the evening when vessel traffic is low...

15 CFR 970.205 - Vessel safety.

Science.gov (United States)

2010-01-01

... 15 Commerce and Foreign Trade 3 2010-01-01 2010-01-01 false Vessel safety. 970.205 Section 970.205... safety. In order to provide a basis for the necessary determinations with respect to the safety of life... Safety of Life at Sea, 1974 (SOLAS 74) possesses current valid SOLAS 74 certificates; (2) That any...

Guidelines for pressure vessel safety assessment

Science.gov (United States)

Yukawa, S.

1990-04-01

A technical overview and information on metallic pressure containment vessels and tanks is given. The intent is to provide Occupational Safety and Health Administration (OSHA) personnel and other persons with information to assist in the evaluation of the safety of operating pressure vessels and low pressure storage tanks. The scope is limited to general industrial application vessels and tanks constructed of carbon or low alloy steels and used at temperatures between -75 and 315 C (-100 and 600 F). Information on design codes, materials, fabrication processes, inspection and testing applicable to the vessels and tanks are presented. The majority of the vessels and tanks are made to the rules and requirements of ASME Code Section VIII or API Standard 620. The causes of deterioration and damage in operation are described and methods and capabilities of detecting serious damage and cracking are discussed. Guidelines and recommendations formulated by various groups to inspect for the damages being found and to mitigate the causes and effects of the problems are presented.

Modal-split effects of climate change: The effects of low water levels on the competitive position of inland waterway transport in the River Rhine area

NARCIS (Netherlands)

Jonkeren, O.E.; Jourquin, B.; Rietveld, P.

2011-01-01

Future climate change is expected to affect inland waterway transport in most main natural waterways in Europe. For the river Rhine it is expected that, in summer, more and longer periods with low water levels will occur. In periods of low water levels inland waterway vessels have to reduce their

Andean waterways

DEFF Research Database (Denmark)

Rasmussen, Mattias Borg

Andean Waterways explores the politics of natural resource use in the Peruvian Andes in the context of climate change and neoliberal expansion. It does so through careful ethnographic analysis of the constitution of waterways, illustrating how water becomes entangled in a variety of political......, social, and cultural concerns. Set in the highland town of Recuay in Ancash, the book traces the ways in which water affects political and ecological relations as glaciers recede. By looking at the shared waterways of four villages located in the foothills of Cordillera Blanca, it addresses pertinent...

Preliminary Performance Analysis Program Development for Safety System with Safeguard Vessel

International Nuclear Information System (INIS)

Kang, Han-Ok; Lee, Jun; Park, Cheon-Tae; Yoon, Ju-Hyeon; Park, Keun-Bae

2007-01-01

SMART is an advanced modular integral type pressurized water reactor for a seawater desalination and an electricity production. Major components of the reactor coolant system such as the pressurizer, Reactor Coolant Pump (RCP), and steam generators are located inside the reactor vessel. The SMART can fundamentally eliminate the possibility of large break loss of coolant accidents (LBLOCAs), improve the natural circulation capability, and better accommodate and thus enhance a resistance to a wide range of transients and accidents. The safety goals of the SMART are enhanced through highly reliable safety systems such as the passive residual heat removal system (PRHRS) and the safeguard vessel coupled with the passive safety injection feature. The safeguard vessel is a steel-made, leak-tight pressure vessel housing the RPV, SIT, and the associated valves and pipelines. A primary function of the safeguard vessel is to confine any radioactive release from the primary circuit within the vessel under DBAs related to loss of the integrity of the primary system. A preliminary performance analysis program for a safety system using the safeguard vessel is developed in this study. The developed program is composed of several subroutines for the reactor coolant system, passive safety injection system, safeguard vessel including the pressure suppression pool, and PRHRS. A small break loss of coolant accident at the upper part of a reactor is analyzed and the results are discussed

78 FR 26799 - Waterway Suitability Assessment for Expansion of Liquefied Gas Terminals; Beaumont, TX

Science.gov (United States)

2013-05-08

... Partnership LP has submitted a Letter of Intent and a Waterway Suitability Assessment to the Coast Guard.... Martin Operating Partnership's Stanolind Cut Terminal in Beaumont, Texas submitted an LOI on January 25... the physical nature of the affected waterway and issues of safety and security associated with LNG or...

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

Andean waterways

DEFF Research Database (Denmark)

Rasmussen, Mattias Borg

, social, and cultural concerns. Set in the highland town of Recuay in Ancash, the book traces the ways in which water affects political and ecological relations as glaciers recede. By looking at the shared waterways of four villages located in the foothills of Cordillera Blanca, it addresses pertinent......Andean Waterways explores the politics of natural resource use in the Peruvian Andes in the context of climate change and neoliberal expansion. It does so through careful ethnographic analysis of the constitution of waterways, illustrating how water becomes entangled in a variety of political...... questions concerning water governance and rural lives....

Model of passage of vessels through the waterway section forming the schedule in the course of its implementation

Directory of Open Access Journals (Sweden)

S. N. Plotnikov

2017-01-01

Full Text Available The model of passage of vessels through a section of the waterway is considered, which independently determines the order of passage of vessels with limited capacity of sections of the track. Such a model will consist of a number of standard algorithmic networks. When composing the schedule in the model, the following preference rules were used: first-come-first-served (that is, if the ship occupied the workplace, this decision is not canceled; The rule of the shortest operation; For the swamps the priority of vessels going downstream (the direction of flow from the source to the drain. An algorithmic network that implements the search for an acceptable schedule must implement the following for conflicting operations: the operation that has started is not interrupted; If several operations simultaneously claim for one workplace (port, reach and their number is greater than its throughput, then the conflict resolution is carried out in accordance with predefined preference rules or based on the user's decision; If the operation is waiting for the release of the workplace, it does not occupy the resource; The resource is returned immediately after the operation is completed. The considered design of algorithmic networks allows to resolve the conflict, with a simultaneous resource request, to take the resource once, remember that it was received and return it after the end of the operation, then the resource receives a contra-controlling operation for execution. However, the use of this design introduces redundancy into the model, even if it is used only for conflicting operations. The model is presented in the language of algorithmic networks and is implemented in the system of modeling automation KOGNITRON.

The Coast Guard Proceedings of the Marine Safety and Security Council. Volume 72, Number 4, Winter 2015-2016

Science.gov (United States)

2016-04-13

National Drug Threat Assessment reported that the annual economic cost of controlled prescription drug non-medical use (use beyond that of...scenario: On a clear Tuesday eve- ning, an outbound small coastal freight ship collides with an inbound offshore supply vessel while navigating a bend...NMCProceedings@uscg.mil Website: www.uscg.mil/proceedings Safety Management Systems Waterways: America’s Economic Engine Recreational Boating Safety i

76 FR 13422 - Lower Mississippi River Waterway Safety Advisory Committee; Vacancies

Science.gov (United States)

2011-03-11

... the Lower Mississippi River and its connecting navigable waterways, including the Gulf of Mexico. The... receive no salary, reimbursement of travel expenses, or other compensation from the Federal Government...

76 FR 60076 - Lower Mississippi River Waterway Safety Advisory Committee; Vacancies

Science.gov (United States)

2011-09-28

... navigable waterways, including the Gulf of Mexico. 1. Five members representing River Port Authorities... consecutive terms. All members serve at their own expense and receive no salary reimbursement of travel...

Safety of steel vessel Magnox pressure circuits

International Nuclear Information System (INIS)

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

1991-01-01

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

77 FR 35271 - Safety Zone; NOAA Vessel Rueben Lasker Launch, Marinette, WI

Science.gov (United States)

2012-06-13

...-AA00 Safety Zone; NOAA Vessel Rueben Lasker Launch, Marinette, WI AGENCY: Coast Guard, DHS. ACTION... during the launching of the NOAA vessel, Rueben Lasker, on June 16, 2012. This temporary safety zone is... preceding paragraph, a 30 day notice period would also be impractical. B. Basis and Purpose The NOAA vessel...

Reasons and remedies of inland passenger vessels accidents in Bangladesh

Science.gov (United States)

Rashid, Cdr Kaosar; Islam, Muhammad Rabiul

2017-12-01

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

78 FR 76751 - Safety Zone; Vessel Launch; Menominee River; Marinette, WI

Science.gov (United States)

2013-12-19

...-AA00 Safety Zone; Vessel Launch; Menominee River; Marinette, WI AGENCY: Coast Guard, DHS. ACTION: Temporary final rule. SUMMARY: The Coast Guard is establishing a temporary safety zone on the Menominee River in Marinette, Wisconsin. This zone is intended to restrict vessels from a portion of the Menominee...

77 FR 60042 - Safety Zone; Research Vessel SIKULIAQ Launch, Marinette, WI

Science.gov (United States)

2012-10-02

...: Temporary final rule. SUMMARY: The Coast Guard is establishing a temporary safety zone on the Menominee River in Marinette Wisconsin. This zone is intended to restrict vessels from a portion of Menominee River during the launching of the Research vessel SIKULIAQ, on October 13th, 2012. This temporary safety...

33 CFR 207.100 - Inland waterway from Delaware River to Chesapeake Bay, Del. and Md. (Chesapeake and Delaware...

Science.gov (United States)

2010-07-01

... enter or pass through any part of the waterway will be contingent on the vessel's having adequate... facilities are of limited capacity, and permission to occupy them for periods exceeding 24 hours must be...

Inland Waterway Mile Markers

Data.gov (United States)

Department of Homeland Security — Contains 11,201 sequential mile positions of navigable inland waterways and the Gulf Intracoastal Waterway. Records include eight fields; id, latitude, longitude,...

Safety of nuclear pressure vessels and its regulatory aspects in France

Energy Technology Data Exchange (ETDEWEB)

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

1979-01-01

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

77 FR 3115 - Safety Zone; Grain-Shipment Vessels, Columbia and Snake Rivers

Science.gov (United States)

2012-01-23

...-AA00 Safety Zone; Grain-Shipment Vessels, Columbia and Snake Rivers AGENCY: Coast Guard, DHS. ACTION... Terminal, Longview, WA, while they are located on the Columbia and Snake Rivers. This safety zone extends... on the Columbia and Snake rivers when vessels begin arriving at EGT, Longview, WA. Under 5 U.S.C. 553...

Initiation and arrest - two approaches to pressure vessel safety

International Nuclear Information System (INIS)

Brumovsky, M.; Filip, R.; Stepanek, S.

1976-01-01

The safety analysis is described of the reactor pressure vessel related to brittle fracture based on the fracture mechanics theory using two different approximations, i.e., the Crack Arrest Temperature (CAT) or Nil Ductility Temperature (NDT), and fracture toughness. The variation of CAT with stress was determined for different steel specimens of 120 to 200 mm in thickness. A diagram is shown of CAT variation with stress allowing the determination of crack arrest temperature for all types of commonly used steels independently of the NDT initial value. The diagram also shows that the difference between fracture transition elastic (FTE) and NDT depends on the type of material and determines the value of the ΔTsub(sigma) factor typical of the safety coefficient. The so-called fracture toughness reference value Ksub(IR) is recommended for the computation of pressure vessel criticality. Also shown is a defect analysis diagram which may be used for the calculation of pressure vessel safety prior to and during operation and which may also be used in making the decision on what crack sizes are critical, what cracks may be arrested and what cracks are likely to expand. The diagram is also important for the fact that it is material-independent and may be employed for the estimates of pre-operational and operational inspections and for pressure vessel life prediction. It is generally applicable to materials of greater thickness in the region where the validity of linear elastic fracture mechanics is guaranteed. (J.P.)

33 CFR 162.65 - All waterways tributary to the Atlantic Ocean south of Chesapeake Bay and all waterways tributary...

Science.gov (United States)

2010-07-01

... Atlantic Ocean south of Chesapeake Bay and all waterways tributary to the Gulf of Mexico east and south of... All waterways tributary to the Atlantic Ocean south of Chesapeake Bay and all waterways tributary to..., which are tributary to or connected by other waterways with the Atlantic Ocean south of Chesapeake Bay...

Development of Safety Review Guide for the Periodic Safety Review of Reactor Vessel Internals

International Nuclear Information System (INIS)

Park, Jeongsoon; Ko, Hanok; Kim, Seonjae; Jhung, Myungjo

2013-01-01

Aging management of the reactor vessel internals (RVIs) is one of the important issues for long-term operation of nuclear power plants (NPPs). Safety review on the assessment and management of the RVI aging is conducted through the process of a periodic safety review (PSR). The regulatory body should check that reactor facilities sustain safety functions in light of degradation due to aging and that the operator of a nuclear power reactor establishes and implements management program to deal with degradation due to aging in order to guarantee the safety functions and the safety margin as a result of PSR. KINS(Korea Institute of Nuclear Safety) has utilized safety review guides (SRG) which provide guidance to KINS staffs in performing safety reviews in order to assure the quality and uniformity of staff safety reviews. The KINS SRGs for the continued operation of pressurized water reactors (PWRs) published in 2006 contain areas of review regarding aging management of RVIs in chapter 2 (III.2.15, Appendix 2.0.1). However unlike the SRGs for the continued operation, KINS has not officially published the SRGs for the PSR of PWRs, but published them as a form of the research report. In addition to that, the report provides almost same review procedures for aging assessment and management of RVIs with the ones provided in the SRGs for the continued operation, it cannot provide review guidance specific to PSRs. Therefore, a PSR safety review guide should be developed for RVIs in PWRs. In this study, a draft PSR safety review guide for reactor vessel internals in PWRs is developed and provided. In this paper, a draft PSR safety review guide for reactor vessel internals (PSR SRG-RVIs) in PWRs is introduced and main contents of the draft are provided. However, since the PSR safety review guides for areas other than RVIs in the pressurized water reactors (PWRs) are expected to be developed in the near future, the draft PSR SRG-RVIs should be revisited to be compatible with

National Waterway Network (line), Geographic WGS84, BTS (2006) [usace_nav_waterway_lin_BTS_2006

Data.gov (United States)

Louisiana Geographic Information Center — The National Waterway Network is a comprehensive network database of the nation's navigable waterways. The data set covers the 48 contiguous states plus the District...

National Waterway Network (node), Geographic WGS84, BTS (2006) [usace_nav_waterway_nod_BTS_2006

Data.gov (United States)

Louisiana Geographic Information Center — The National Waterway Network is a comprehensive network database of the nation's navigable waterways. The data set covers the 48 contiguous states plus the District...

Integral reactor vessel related to power reactor safety

International Nuclear Information System (INIS)

Widart, J.; Scailteur, A.

1978-01-01

Integral design applied to PWR pressure vessels allows to reach a high level of safety because: 1) it presents a better balance of the material in the geometry, resulting in an improved stress level (mainly faulted condition loadings); 2) location and geometry of the welds are designed in order to get a very sound pressure boundary of the upper part of the vessel; 3) the new location and geometry of the welds allow an easy ISI in such a way that ambiguity surrounding defect size or locaton is practically suppressed. (author)

Digital waterway construction based on inland electronic navigation chart

Science.gov (United States)

Wang, Xue; Pan, Junfeng; Zhu, Weiwei

2015-12-01

With advantages of large capacity, long distance, low energy consumption, low cost, less land occupation and light pollution, inland waterway transportation becomes one of the most important constituents of the comprehensive transportation system and comprehensive water resources utilization in China. As one of "three elements" of navigation, waterway is the important basis for the development of water transportation and plays a key supporting role in shipping economic. The paper discuss how to realize the informatization and digitization of waterway management based on constructing an integrated system of standard inland electronic navigation chart production, waterway maintenance, navigation mark remote sensing and control, ship dynamic management, and water level remote sensing and report, which can also be the foundation of the intelligent waterway construction. Digital waterway construction is an information project and also has a practical meaning for waterway. It can not only meet the growing high assurance and security requirements for waterway, but also play a significant advantage in improving transport efficiency, reducing costs, promoting energy conservation and so on. This study lays a solid foundation on realizing intelligent waterway and building a smooth, efficient, safe, green modern inland waterway system, and must be considered as an unavoidable problem for the coordinated development between "low carbon" transportation and social economic.

Safety margins of PWR irradiated vessels - The Chooz A issue

Energy Technology Data Exchange (ETDEWEB)

Hedin, F; Barthelet, B [Electricite de France (EDF), 75 - Paris (France); Guilleret, J C

1988-12-31

In 1986, some irradiated specimen of CHOOZ A (SENA) vessel showed a significant excess of {delta} RTNDT to former previsions. The lack of data on one of the two irradiated shells, and discrepancies between dosimeters results and available previous fluence calculations whose accuracy was questionable, cause the safety authorities to require an important complementary work program before putting again the plant on the grid after 1987 fuel reloading. These works are presented and discussed. They lead to a state that a conservative to day value of the vessel RTNDT is 64 degrees Celsius and that there is no underclad defect in the vessel wall and welds. Then the plant was allowed to restart with certitude that vessel irradiation will not impair its lifetime. (author). 4 refs.

An insight into the maritime accident characteristics in Bangladesh

Science.gov (United States)

Uddin, Md. Imran; Awal, Zobair Ibn

2018-03-01

The inland waterway plays a very important role in the transportation system of Bangladesh. But, due to severe deficiencies of the safety practices, a lot of accidentll take place almost every year in the inland waterways that cause considerable loss of human lives and assets. The inland waterway accidentll in Bangladasb during 2005 to 2015 were analysed statistically in the present study. It has been found that the leading mode of accidents is collision among vessels and fatal injury comprises considerably a higher proportion of total casnalties. The study also revealed that cargo vessels and passenger vessels encounter more accidents in comparison to other types of vessels. It was also observed that during fair weather and good visibility condition significant proportion of accidents have taken place. Based on this study several recommendstions bsve been put forward for improving ssfety in the inlsnd waterways ofBanglsdesh.

An insight into the maritime accident characteristics in Bangladesh

Science.gov (United States)

Uddin, Md. Imran; Awal, Zobair Ibn

2017-12-01

The inland waterway plays a very important role in the transportation system of Bangladesh. But, due to severe deficiencies of the safety practices, a lot of accidentll take place almost every year in the inland waterways that cause considerable loss of human lives and assets. The inland waterway accidentll in Bangladasb during 2005 to 2015 were analysed statistically in the present study. It has been found that the leading mode of accidents is collision among vessels and fatal injury comprises considerably a higher proportion of total casnalties. The study also revealed that cargo vessels and passenger vessels encounter more accidents in comparison to other types of vessels. It was also observed that during fair weather and good visibility condition significant proportion of accidents have taken place. Based on this study several recommendstions bsve been put forward for improving ssfety in the inlsnd waterways ofBanglsdesh.

Probabilistic Assessment of the Design and Safety of HSLA-100 Steel Confinement Vessels

Energy Technology Data Exchange (ETDEWEB)

R.M. Dolin

2003-03-03

This probabilistic approach for assessing the design and safety of the HSLA-100 steel confinement vessel used for a DynEx test involved the probability of failure for several scenarios, in which a fragment may penetrate the vessel. The samples involve vessel thicknesses of 1 inch, 2 inches, and 5.25 inches--the combined thicknesses of the 2 inch containment vessel and the 3.25 inch safety vessel. Two simulation approaches were used for each scenario to assess the probability of failure. The Likelihood of Occurrence method simultaneously models all likely fragment events of a test, for which the net probability of failure is the sum of all the fragment events. The Stochastic Sampling method determines the probability of a fragment perforation on the basis of a logical model and takes the overall probability that an experiment results in failure as the maximum probability for any fragment event. With margin and safety assessments taken into account, it was concluded that the one and two inch thicknesses by themselves are inadequate for containing a DynEx test. The 5.25 inch thickness was determined to be safe by the Likelihood of Occurrence method and nearly adequate by the Stochastic Sampling simulation.

Analysis Of Inland Waterway Transport For Container Shipping: Cikarang To Port Of TanjungPriok

Science.gov (United States)

Achmadi, T.; Nur, H. I.; Rahmadhon, L. R.

2018-03-01

Industry's development which is in the center of Cikarang's industrial estate causes a considerable increase from 7% to 13% of container's flow from and to Port of Tanjung Priok per year. Therefore, those obstacles rise the number of traffic congestion and transport cost. This research aims to analyze the potential alternative of transportation in order to transport containers at the route of Tanjung Priok to Cikarang utilizing Inland Waterways Transport through Cikarang Bekasi Laut (CBL) river. This research will be conducted by comparing component of total logistic cost that emerging caused by container trucks and vessels. Self Propelled Container Barge (SPCB) is a pointed alternative transportation in which it is used to transport containers through the waterways. The result of analysis obtained that the capacity of Cikarang Bekasi Laut river is 18,558 roundtrip per year. Furthermore, the collaboration of 3 SPCB operations, as well payload 32 TEUS can decrease the amount of road traffic congestion/density of Cikarang-Port of Tanjung Priok as much as 18.6%. The cost of containers shipping per unit transported by truck is IDR 2.2 Million per TEUs, whereas containers shipping transported by Inland Waterways cost only IDR 1.8 Million per TEUs.

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...

33 CFR 161.4 - Requirement to carry the rules.

Science.gov (United States)

2010-07-01

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

Safety of light-water reactor pressure vessels against brittle fracture

International Nuclear Information System (INIS)

Brumovsky, M.

1979-01-01

The results are surveyed of research by SKODA Trust into brittle failure resistance of materials for WWER type reactor pressure vessels and into pressure vessel operating safety. Conditions are discussed in detail decisive for initiation, propagation and arrest of brittle fracture. The tests on the Cr-Mo-V type steel showed high resistance of the steel to the formation and the propagation of brittle fracture. They also confirmed the high operating reliability and the required service life of the steel. (B.S.)

78 FR 33224 - Safety Zone; Grain-Shipment and Grain-Shipment Assist Vessels, Columbia and Willamette Rivers

Science.gov (United States)

2013-06-04

... 1625-AA00 Safety Zone; Grain-Shipment and Grain-Shipment Assist Vessels, Columbia and Willamette Rivers... Guard is establishing a temporary safety zone around all inbound and outbound grain-shipment and grain-shipment assist vessels involved in commerce with the Columbia Grain facility on the Willamette River in...

78 FR 57261 - Safety Zone; Grain-Shipment and Grain-Shipment Assist Vessels, Columbia and Willamette Rivers

Science.gov (United States)

2013-09-18

... 1625-AA00 Safety Zone; Grain-Shipment and Grain-Shipment Assist Vessels, Columbia and Willamette Rivers... temporary safety zone around all inbound and outbound grain-shipment and grain-shipment assist vessels involved in commerce with the Columbia Grain facility on the Willamette River in Portland, OR, the United...

76 FR 27897 - Security and Safety Zone Regulations, Large Passenger Vessel Protection, Captain of the Port...

Science.gov (United States)

2011-05-13

... DEPARTMENT OF HOMELAND SECURITY Coast Guard 33 CFR Part 165 [Docket No. USCG-2011-0342] Security and Safety Zone Regulations, Large Passenger Vessel Protection, Captain of the Port Columbia River... will enforce the security and safety zone in 33 CFR 165.1318 for large passenger vessels operating in...

CLIMATE CHANGE AND THE RISK MANAGEMENT IN SERBIAN INLAND WATERWAY TRANSPORT

Directory of Open Access Journals (Sweden)

Todor Backalic

2012-12-01

Full Text Available Climate changes require special attention and global monitoring. Impacts on nature and industry are increasingly evident. Water transport could be the backbone of the future European combined transport system. Unfortunately, risks in water transport are perhaps an under researched area and consequently, this article outlines a rationale, why it is necessary to develop competence about risk in water transport. The development of the cargo transport in river traffic depends directly on technical-exploitative characteristics of the network of inland waterways. Research of navigational abilities of inland waterways always comes before building ships or making a transport schedule. Navigation characteristics of rivers have to be determined as precise as possible, especially from the aspect of determination the possible draught of vessels. River transport due to their operational business and environmental conditions faced with several risks having different degrees of consequences. Current risk assessment methods for water transport just consider some dramatic events. We present a new method for the assessment of risk and vulnerability of water transport where river depth represents a crucial part. The analysis of water level changes on Serbian rivers during the last sixty years was done.

Supporting inland waterway transport on German waterways by operational forecasting services - water-levels, discharges, river ice

Science.gov (United States)

Meißner, Dennis; Klein, Bastian; Ionita, Monica; Hemri, Stephan; Rademacher, Silke

2017-04-01

Inland waterway transport (IWT) is an important commercial sector significantly vulnerable to hydrological impacts. River ice and floods limit the availability of the waterway network and may cause considerable damages to waterway infrastructure. Low flows significantly affect IWT's operation efficiency usually several months a year due to the close correlation of (low) water levels / water depths and (high) transport costs. Therefore "navigation-related" hydrological forecasts focussing on the specific requirements of water-bound transport (relevant forecast locations, target parameters, skill characteristics etc.) play a major role in order to mitigate IWT's vulnerability to hydro-meteorological impacts. In light of continuing transport growth within the European Union, hydrological forecasts for the waterways are essential to stimulate the use of the free capacity IWT still offers more consequently. An overview of the current operational and pre-operational forecasting systems for the German waterways predicting water levels, discharges and river ice thickness on various time-scales will be presented. While short-term (deterministic) forecasts have a long tradition in navigation-related forecasting, (probabilistic) forecasting services offering extended lead-times are not yet well-established and are still subject to current research and development activities (e.g. within the EU-projects EUPORIAS and IMPREX). The focus is on improving technical aspects as well as on exploring adequate ways of disseminating and communicating probabilistic forecast information. For the German stretch of the River Rhine, one of the most frequented inland waterways worldwide, the existing deterministic forecast scheme has been extended by ensemble forecasts combined with statistical post-processing modules applying EMOS (Ensemble Model Output Statistics) and ECC (Ensemble Copula Coupling) in order to generate water level predictions up to 10 days and to estimate its predictive

The analysis of inland water transport on technically developed Polish section of the E70 waterway using GIS tools in the years 2005-2014

Science.gov (United States)

Rabant, Hubert; Szatten, Dawid; Nadolny, Grzegorz

2017-11-01

The article presents the characteristics of changes in the spatial extent of transport on the hydrotechnically developed section of the E70 waterway in Poland using methods and tools of geographic information systems (GIS). The results of the analyzes show the conditions for vessel traffic, their type and volatility in the years 2005-2014. The methods made it possible to analyze the spatial determinants of navigation. The obtained results were referred to the current state and prospects for development of Polish waterways and indicated that the applied tools have a great application role in the research on their logistics and development.

Safety analysis of nuclear containment vessels subjected to strong earthquakes and subsequent tsunamis

Energy Technology Data Exchange (ETDEWEB)

Lin, Feng; Li, Hong Zhi [Dept. Structural Engineering, Tongji University, Shanghai (China)

2017-08-15

Nuclear power plants under expansion and under construction in China are mostly located in coastal areas, which means they are at risk of suffering strong earthquakes and subsequent tsunamis. This paper presents a safety analysis for a new reinforced concrete containment vessel in such events. A finite element method-based model was built, verified, and first used to understand the seismic performance of the containment vessel under earthquakes with increased intensities. Then, the model was used to assess the safety performance of the containment vessel subject to an earthquake with peak ground acceleration (PGA) of 0.56g and subsequent tsunamis with increased inundation depths, similar to the 2011 Great East earthquake and tsunami in Japan. Results indicated that the containment vessel reached Limit State I (concrete cracking) and Limit State II (concrete crushing) when the PGAs were in a range of 0.8–1.1g and 1.2–1.7g, respectively. The containment vessel reached Limit State I with a tsunami inundation depth of 10 m after suffering an earthquake with a PGA of 0.56g. A site-specific hazard assessment was conducted to consider the likelihood of tsunami sources.

Safeguarding the nuclear safety of WWER-440 reactor pressure vessels at SKODA Plzen

International Nuclear Information System (INIS)

Hrbek, Z.

1986-01-01

The approach is described of the SKODA enterprise to safety assurance and to providing the reliability of WWER-440 reactor pressure vessels. The philosophy is analyzed of in-service inspection and determination of the residual service life of pressure vessels. This follows up on the so-called conception of basic safety whose main aim is to preclude failures at production stage by the selection of suitable material, namely by optimizing the choice of raw materials, of metallurgical procedures such as will lead to high purity of the pressure vessel material, by introducing multiple inspection in production, reducing the sensitivity of materials to technological operations, and by high-quality welds. The quality of in-service inspections is given by the use of technical diagnostic instruments of peak quality and of modern methods of nondestructive materials testing. The instruments and methods used are described. It is stated that the experience gained with in-service inspection will make it possible to draw up operating regulations and safety criteria for nuclear installations and own inspection regulations, this with regard to technical and economic factors. (Z.M.)

78 FR 68995 - Safety Zone: Vessel Removal From the Oakland Estuary, Alameda, CA

Science.gov (United States)

2013-11-18

...-AA00 Safety Zone: Vessel Removal From the Oakland Estuary, Alameda, CA AGENCY: Coast Guard, DHS. ACTION... waters of the Oakland Estuary just north of the Park Street Bridge in Alameda, CA in support of the Oakland Estuary Closure for the Vessel Removal Project on November 4, 2013 through November 22, 2013. This...

75 FR 11870 - Inland Waterways Users Board

Science.gov (United States)

2010-03-12

... DEPARTMENT OF DEFENSE Department of the Army; Corps of Engineers Inland Waterways Users Board AGENCY: Department of the Army, U.S. Army Corps of Engineers, DoD. ACTION: Notice of open meeting... is made of the forthcoming meeting. Name of Committee: Inland Waterways Users Board (Board). Date...

78 FR 44934 - Inland Waterways Users Board

Science.gov (United States)

2013-07-25

... DEPARTMENT OF DEFENSE Department of the Army; Corps of Engineers Inland Waterways Users Board AGENCY: Department of the Army, U.S. Army Corps of Engineers, DoD. ACTION: Notice of open meeting... is made of the forthcoming meeting. Name of Committee: Inland Waterways Users Board (Board). Date...

75 FR 57264 - Inland Waterways Users Board

Science.gov (United States)

2010-09-20

... DEPARTMENT OF DEFENSE Department of the Army; Corps of Engineers Inland Waterways Users Board AGENCY: Department of the Army, U.S. Army Corps of Engineers, DoD. ACTION: Notice of open meeting... is made of the forthcoming meeting. Name of Committee: Inland Waterways Users Board (Board). Date...

76 FR 11216 - Inland Waterways Users Board

Science.gov (United States)

2011-03-01

... DEPARTMENT OF DEFENSE Department of the Army; Corps of Engineers Inland Waterways Users Board AGENCY: Department of the Army, U.S. Army Corps of Engineers, DoD. ACTION: Notice of open meeting... is made of the forthcoming meeting. Name of Committee: Inland Waterways Users Board (Board). Date...

78 FR 72070 - Inland Waterways Users Board

Science.gov (United States)

2013-12-02

... DEPARTMENT OF DEFENSE Department of the Army; Corps of Engineers Inland Waterways Users Board AGENCY: Department of the Army, U.S. Army Corps of Engineers, DoD. ACTION: Notice of open meeting... is made of the forthcoming meeting. Name of Committee: Inland Waterways Users Board (Board). Date...

77 FR 69447 - Inland Waterways Users Board

Science.gov (United States)

2012-11-19

... DEPARTMENT OF DEFENSE Department of the Army; Corps of Engineers Inland Waterways Users Board AGENCY: Department of the Army, U.S. Army Corps of Engineers, DoD. ACTION: Notice of open meeting... is made of the forthcoming meeting. Name of Committee: Inland Waterways Users Board (Board). Date...

Safety analysis of nuclear containment vessels subjected to strong earthquakes and subsequent tsunamis

Directory of Open Access Journals (Sweden)

Feng Lin

2017-08-01

Full Text Available Nuclear power plants under expansion and under construction in China are mostly located in coastal areas, which means they are at risk of suffering strong earthquakes and subsequent tsunamis. This paper presents a safety analysis for a new reinforced concrete containment vessel in such events. A finite element method-based model was built, verified, and first used to understand the seismic performance of the containment vessel under earthquakes with increased intensities. Then, the model was used to assess the safety performance of the containment vessel subject to an earthquake with peak ground acceleration (PGA of 0.56g and subsequent tsunamis with increased inundation depths, similar to the 2011 Great East earthquake and tsunami in Japan. Results indicated that the containment vessel reached Limit State I (concrete cracking and Limit State II (concrete crushing when the PGAs were in a range of 0.8–1.1g and 1.2–1.7g, respectively. The containment vessel reached Limit State I with a tsunami inundation depth of 10 m after suffering an earthquake with a PGA of 0.56g. A site-specific hazard assessment was conducted to consider the likelihood of tsunami sources.

76 FR 31350 - Cruise Vessel Safety and Security Act of 2010, Available Technology

Science.gov (United States)

2011-05-31

... DEPARTMENT OF HOMELAND SECURITY Coast Guard [Docket No. USCG-2011-0357] Cruise Vessel Safety and Security Act of 2010, Available Technology AGENCY: Coast Guard, DHS. ACTION: Notice of request for comments... Security and Safety Act of 2010(CVSSA), specifically related to video recording and overboard detection...

Safety assessment of a multicavity prestressed concrete reactor vessel with hot liner

Energy Technology Data Exchange (ETDEWEB)

Lafitte, R.; Marchand, J. D. [Bonnard et Gardel, Ingenieurs-Conseil, Lausanne (Switzerland)

1981-01-15

The prestressed concrete reactor vessel of the high temperature reactor with helium turbine project differs from those realized up to this day by the important number of cavities, by the different cavity pressures and by a liner in contact with hot gas. For the cases of operating conditions, the computations can be based on an identical pressure in all the cavities. The overdimensioning of the vessel which results is not a determining factor at this stage of the project. The possible loss of leaktightness of the liner can introduce gas pressure into the walls of the vessel. The great thickness of the walls makes it impossible to withstand the resulting forces with prestressing in offering sufficient safety factor against collapse. It is thus important to design a drainage network largely dimensioned. The warm liner appears at this stage of the project too highly stressed by fatigue at the singularity points (ducts between cavities, angles). A solution is proposed which limits the variations of thermal stresses by using a steel with low coefficient of thermal expansion. The cavity closures, which are numerous and some with large dimensions are an important aspect of the vessel safety. A solution of reinforced concrete shell with independent liner is proposed.

Safety assessment of a multicavity prestressed concrete reactor vessel with hot liner

International Nuclear Information System (INIS)

Lafitte, R.; Marchand, J.D.

1981-01-01

The prestressed concrete reactor vessel of the high temperature reactor with helium turbine project differs from those realized up to this day by the important number of cavities, by the different cavity pressures and by a liner in contact with hot gas. For the cases of operating conditions, the computations can be based on an identical pressure in all the cavities. The overdimensioning of the vessel which results is not a determining factor at this stage of the project. The possible loss of leaktightness of the liner can introduce gas pressure into the walls of the vessel. The great thickness of the walls makes it impossible to withstand the resulting forces with prestressing in offering sufficient safety factor against collapse. It is thus important to design a drainage network largely dimensioned. The warm liner appears at this stage of the project too highly stressed by fatigue at the singularity points (ducts between cavities, angles). A solution is proposed which limits the variations of thermal stresses by using a steel with low coefficient of thermal expansion. The cavity closures, which are numerous and some with large dimensions are an important aspect of the vessel safety. A solution of reinforced concrete shell with independent liner is proposed

77 FR 22769 - Amendment to the Inland Waterways Users Board

Science.gov (United States)

2012-04-17

... DEPARTMENT OF DEFENSE Office of the Secretary Amendment to the Inland Waterways Users Board AGENCY... charter for the Inland Waterways Users Board (hereafter referred to as ``the Board''). The Board is... shall invite primary commercial users and shippers of the inland and intracoastal waterways to serve on...

75 FR 14609 - Commercial Fishing Industry Vessel Safety Advisory Committee; Vacancies

Science.gov (United States)

2010-03-26

... which Chapter 45 of Title 46, U.S.C. applies and persons representing the marine insurance industry... Industry Vessel Safety Advisory Committee; Vacancies AGENCY: Coast Guard, DHS. ACTION: Request for applications. SUMMARY: The Coast Guard seeks applications for membership on the Commercial Fishing Industry...

76 FR 14279 - Safety Zone; Todd Pacific Shipyards Vessel Roll-Out, West Duwamish Waterway, Seattle, WA

Science.gov (United States)

2011-03-16

... from entering or remaining in the safety zone unless authorized by the Captain of the Port or a... published. In addition, given the dangers involved with a large slow moving dry dock maneuvering close to... with a large slow moving dry dock maneuvering close to the shore, delaying the effective date of this...

Development of a monthly to seasonal forecast framework tailored to inland waterway transport in central Europe

Science.gov (United States)

Meißner, Dennis; Klein, Bastian; Ionita, Monica

2017-12-01

Traditionally, navigation-related forecasts in central Europe cover short- to medium-range lead times linked to the travel times of vessels to pass the main waterway bottlenecks leaving the loading ports. Without doubt, this aspect is still essential for navigational users, but in light of the growing political intention to use the free capacity of the inland waterway transport in Europe, additional lead time supporting strategic decisions is more and more in demand. However, no such predictions offering extended lead times of several weeks up to several months currently exist for considerable parts of the European waterway network. This paper describes the set-up of a monthly to seasonal forecasting system for the German stretches of the international waterways of the Rhine, Danube and Elbe rivers. Two competitive forecast approaches have been implemented: the dynamical set-up forces a hydrological model with post-processed outputs from ECMWF general circulation model System 4, whereas the statistical approach is based on the empirical relationship (teleconnection) of global oceanic, climate and regional hydro-meteorological data with river flows. The performance of both forecast methods is evaluated in relation to the climatological forecast (ensemble of historical streamflow) and the well-known ensemble streamflow prediction approach (ESP, ensemble based on historical meteorology) using common performance indicators (correlation coefficient; mean absolute error, skill score; mean squared error, skill score; and continuous ranked probability, skill score) and an impact-based evaluation quantifying the potential economic gain. The following four key findings result from this study: (1) as former studies for other regions of central Europe indicate, the accuracy and/or skill of the meteorological forcing used has a larger effect than the quality of initial hydrological conditions for relevant stations along the German waterways. (2) Despite the predictive

Reactor pressure vessels safety and reliability - certainty and uncertainty

International Nuclear Information System (INIS)

O'Neil, R.

1977-01-01

In the paper, it is suggested that the hazard to the population which would result from vessel failure rate of the order of 10 -6 to 10 -7 per vessel year could be acceptable to society on the basis of other natural and man-made risks. The paper considers the problems of demonstrating safety by calculation based on fracture mechanics, and indicates some of the uncertainties, and inconsistencies in the theory, particularly the effect of cracks in locally degraded volumes of material. The phenomenon of crack arrest is considered, and attention is drawn to the uncertainties as indicated at least by some tests. There is need for speedy resolution of this problem. The uncertainties in material properties, heat treatment and residual stresses are considered, and a proposed upper limit for residual defects ('original sin') is proposed. (orig.) [de

Device for removing hydrogen gas from the safety containment vessel of a nuclear reactor

International Nuclear Information System (INIS)

Stiefel, M.

1983-01-01

The safe processing of all concentrations of gas mixtures should be possible with such a device using a thermal recombiner of compact construction. A recombiner consisting of a metal case and diverter sheets situated in it is heated by induction. The incoming pipe for the gas mixture enriched with hydrogen and the outgoing pipe for the gas mixture with low hydrogen content are connected together by a three way valve. The third connection to the safety valve takes the larger port of the gas mixture with low hydrogen content back to the safety containment vessel. Sufficient amount of the gas mixture with low hydrogen content is taken via the three way valve to the safety containment vessel to ensure that the hydrogen content of the gas mixture taken to the recombiner remains below the 4% by volume limit. (orig./PW)

Occupational Safety and Health Conditions Aboard Small- and Medium-Size Fishing Vessels: Differences among Age Groups.

Science.gov (United States)

Zytoon, Mohamed A; Basahel, Abdulrahman M

2017-02-24

Although marine fishing is one of the most hazardous occupations, research on the occupational safety and health (OSH) conditions aboard marine fishing vessels is scarce. For instance, little is known about the working conditions of vulnerable groups such as young and aging fishermen. The objective of the current paper is to study the OSH conditions of young and aging fishermen compared to middle-aged fishermen in the small- and medium-size (SM) marine fishing sector. A cross-sectional study was designed, and 686 fishermen working aboard SM fishing vessels were interviewed to collect information about their safety and health. The associations of physical and psychosocial work conditions with safety and health outcomes, e.g., injuries, illnesses and job satisfaction, are presented. The results of the current study can be utilized in the design of effective accident prevention and OSH training programs for the three age groups and in the regulation of working conditions aboard fishing vessels.

33 CFR 165.20 - Safety zones.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety zones. 165.20 Section 165... WATERWAYS SAFETY REGULATED NAVIGATION AREAS AND LIMITED ACCESS AREAS Safety Zones § 165.20 Safety zones. A Safety Zone is a water area, shore area, or water and shore area to which, for safety or environmental...

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

Science.gov (United States)

2012-07-30

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

Transport and handling of dangerous goods. Training of persons in charge of vehicles or vessels carrying dangerous goods by road or by inland waterways (Dangerous Goods 1979 No.1)

International Nuclear Information System (INIS)

1979-01-01

This Order supplements the Regulations of 15 April 1945 on the transport of dangerous goods by rail, land and inland waterways. It deals with the training of persons in charge of vehicles or boats carrying dangerous goods by road or by inland waterways. It refers in particular to transport of radioactive materials. (NEA) [fr

Autonomous radiation monitoring of small vessels

International Nuclear Information System (INIS)

Ziock, K.P.; Cheriyadat, A.; Fabris, L.; Goddard, J.; Hornback, D.; Karnowski, T.; Kerekes, R.; Newby, J.

2011-01-01

Small private vessels are one avenue by which nuclear materials may be smuggled across international borders. While one can contemplate using the land-based approach of radiation portal monitors on the navigable waterways that lead to many ports, these systems are ill-suited to the problem. In contrast to roadways, where lanes segregate vehicles, and motion is well controlled by inspection booths; channels, inlets, and rivers present chaotic traffic patterns populated by vessels of all sizes. A unique solution to this problem is based on a portal-less portal monitor designed to handle free-flowing traffic on roadways with up to five-traffic lanes. The instrument uses a combination of visible-light and gamma-ray imaging to acquire and link radiation images to individual vehicles. This paper presents the results of a recent test of the system in a maritime setting.

78 FR 55230 - Safety and Environmental Management System Requirements for Vessels on the U.S. Outer Continental...

Science.gov (United States)

2013-09-10

...\\ including the regulation of workplace safety and health.\\2\\ The Coast Guard's regulatory authority extends... 147 [Docket No. USCG-2012-0779] RIN 1625-AC05 Safety and Environmental Management System Requirements... a vessel-specific Safety and Environmental Management System (SEMS) that incorporates the management...

Safety of the pressure vessels of water reactors. Prevention of sudden failure

International Nuclear Information System (INIS)

Petrequin, P.; Barrachin, B.

1975-01-01

From the safety view point the primary circuit is considered as the essential barrier against the diffusion of radioactive products in the event of fuel element failure. The safety of the vessel itself, the failure of which is not accounted for in accident analyses, is based chiefly on a series of preventive measures such as the suitable choice of materials and manufacturing process, compliances with detailed specifications concerning tests and defect tolerances, supervision in service. All these points are examined in detail when the safety analysis is performed. In this context the Service de Recherches Metallurgiques Appliquees assists the Department de Surete Nucleaire in the study of special problems such as the prevention of sudden failure and the characterisation of steels as a function of working conditions, particularly neutron irradiation. The report is thus devoted mainly to the presentation of methods to prevent sudden failure, with special emphasis on the limits of application. Some results obtained at the Service de Recherches Metallurgiques Appliquees on steels typical of those used for water reactor vessels (A533 and A508Cl.3) are given by way of example. Part two concentrates on the role of various factors influencing embrittlement by irradiation [fr

33 CFR 164.33 - Charts and publications.

Science.gov (United States)

2010-07-01

...) PORTS AND WATERWAYS SAFETY NAVIGATION SAFETY REGULATIONS § 164.33 Charts and publications. (a) Each vessel must have the following: (1) Marine charts of the area to be transited, published by the National... tables published by private entities using data provided by the National Ocean Service. (ii) Tidal...

Fracture toughness requirements of reactor vessel material in evaluation of the safety analysis report of nuclear power plants

International Nuclear Information System (INIS)

Widia Lastana Istanto

2011-01-01

Fracture toughness requirements of reactor vessel material that must be met by applicants for nuclear power plants construction permit has been investigated in this paper. The fracture toughness should be described in the Safety Analysis Reports (SARs) document that will be evaluated by the Nuclear Energy Regulatory Agency (BAPETEN). Because BAPETEN does not have a regulations or standards/codes regarding the material used for the reactor vessel, especially in the fracture toughness requirements, then the acceptance criteria that applied to evaluate the fracture toughness of reactor vessel material refers to the regulations/provisions from the countries that have been experienced in the operation of nuclear power plants, such as from the United States, Japan and Korea. Regulations and standards used are 10 CFR Part 50, ASME and ASTM. Fracture toughness of reactor vessel materials are evaluated to ensure compliance of the requirements and provisions of the Regulatory Body and the applicable standards, such as ASME or ASTM, in order to assure a reliability and integrity of the reactor vessels as well as providing an adequate safety margin during the operation, testing, maintenance, and postulated accident conditions over the reactor vessel lifetime. (author)

Decree No. 83/7467 of 24 November 1983 on visits of foreign armed forces vessels in Turkish waters

International Nuclear Information System (INIS)

1983-01-01

This Decree (No. 83/7467) lays down the principles and procedures for the visits of foreign armed forces vessels in Turkish harbours and inland waterways and came into force on 15 December 1983. This new Decree supersedes a similar Decree of 27 December 1978. (NEA) [fr

33 CFR 165.1405 - Regulated Navigation Areas and Security Zones; Designated Escorted Vessels-Philippine Sea and...

Science.gov (United States)

2010-07-01

...) PORTS AND WATERWAYS SAFETY REGULATED NAVIGATION AREAS AND LIMITED ACCESS AREAS Specific Regulated...). (a) Regulated navigation area. The following areas, designated by coordinates referencing World...

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.)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

33 CFR 165.500 - Safety/Security Zones; Chesapeake Bay, Maryland.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety/Security Zones; Chesapeake... HOMELAND SECURITY (CONTINUED) PORTS AND WATERWAYS SAFETY REGULATED NAVIGATION AREAS AND LIMITED ACCESS... Safety/Security Zones; Chesapeake Bay, Maryland. (a) Definitions. (1) Certain Dangerous Cargo (CDC) means...

Bank erosion processes in waterways

NARCIS (Netherlands)

Duro, G.; Uijttewaal, W.S.J.; Kleinhans, M; Crosato, A.

2017-01-01

Waterways serve for several functions besides  transporting goods and people. The ecological  importance of navigable rivers has taken much  attention during recent decades bringing efforts  to improve these natural corridors for fauna and  flora (Boeters et al.,

Margins of the law pertaining to water supplies and waterways

International Nuclear Information System (INIS)

Bickel, C.

1981-01-01

The author examines legal questions coming from points of contact of the law pertaining to water supplies and waterways on the one hand with the Waste Management Law, the Atomic Energy Law and Criminal Law on the other hand. He tries to find ways for solving the practical problems which arise with the execution of the law pertaining to water supplies and waterways. (HSCH) [de

75 FR 51473 - Houston/Galveston Navigation Safety Advisory Committee

Science.gov (United States)

2010-08-20

... DEPARTMENT OF HOMELAND SECURITY Coast Guard [Docket No. USCG-2010-0656] Houston/Galveston... Houston/Galveston Navigation Safety Advisory Committee (HOGANSAC) and its working groups will meet in Houston, Texas, to discuss waterway improvements, aids to navigation, area projects impacting safety on...

Safety Domain Measurement for Vessels in an Overtaking Situation

Directory of Open Access Journals (Sweden)

Hua-Zhi Hsu

2014-12-01

Full Text Available Marine traffic engineering has been pushed to the limits due to a rising demand in the shipping business. Merchant ships are growing dramatically, both in numbers and in size. To keep pace with current developments, automation seems to be one viable option when it comes to keeping ships running with fewer seafarers available. The aim of this paper is to monitor a modern day mariners’ performance while working in a tense situation. The objective is to define the size of the safety domain whilst overtaking a vessel. The approach was to assess the ship's domain area within a 3 nm wide traffic separation scheme by using a ship handling simulator. From the simulation results, an overtaking domain was determined as 1.36 nm long and 0.4 nm wide. Safety domains in real-life situations were experienced on a much smaller scale compared to the previous findings. The working load for this particular operation is expected to be stressful and highly skilled orientated.

78 FR 69803 - Drawbridge Operation Regulations; Gulf Intracoastal Waterway, Venice, FL

Science.gov (United States)

2013-11-21

...-AA09 Drawbridge Operation Regulations; Gulf Intracoastal Waterway, Venice, FL AGENCY: Coast Guard, DHS... that governs the Hatchett Creek (US-41) Twin Bridges, Gulf Intracoastal Waterway mile 56.9, Venice, FL... mile 56.9 at Venice, FL as listed in Sec. 117.287(b), provides the draw of the Hatchett Creek (US-41...

75 FR 30750 - Drawbridge Operation Regulation; Arkansas Waterway, Little Rock, AR

Science.gov (United States)

2010-06-02

...-AA09 Drawbridge Operation Regulation; Arkansas Waterway, Little Rock, AR AGENCY: Coast Guard, DHS... procedures on the Baring Cross Railroad Drawbridge across the Arkansas Waterway at mile 119.6 at Little Rock....123(b) state that the draw of the Baring Cross Railroad Drawbridge, mile 119.6, at Little Rock, AR is...

33 CFR 165.103 - Safety and Security Zones; LPG Vessel Transits in Portland, Maine, Captain of the Port Zone...

Science.gov (United States)

2010-07-01

... within a 500-yard radius of any Liquefied Petroleum Gas (LPG) vessel while it is moored at the LPG... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety and Security Zones; LPG... and Security Zones; LPG Vessel Transits in Portland, Maine, Captain of the Port Zone, Portsmouth...

A regional waterway management system for balancing recreational boating and resource protection.

Science.gov (United States)

Swett, Robert A; Listowski, Charles; Fry, Douglas; Boutelle, Stephen; Fann, David

2009-06-01

Florida's coasts have been transformed over the past three decades as population growth and unprecedented demand for individual shore access to bays and estuaries led to the creation of residential canal developments. Thousands of miles of channels and basins were dredged as a by-product of this urbanization process. The navigable waterways that resulted are now being stressed by increasing boat traffic and canal-side activities. Recognizing their common goal to preserve the recreational and ecological value of southwest Florida waterways, the Florida Department of Environmental Protection, the four-county West Coast Inland Navigation District, and the University of Florida Sea Grant College Program signed a Memorandum of Agreement. The signatories agreed to develop a science-based Regional Waterway Management System (RWMS), which is a new approach to waterway planning and permitting based on carefully mapped channel depths, a census of actual boat populations, and the spatial extent of natural resources. The RWMS provides a comprehensive, regional overview of channel conditions and the geographic distribution and severity of existing impediments to safe navigation and resource protection. RWMS information and analyses result in regional-scale permitting to accommodate water-dependent uses while minimizing environmental impacts and reducing public expenditures. Compared with traditional approaches to waterway management, the science-based RWMS is relatively unbiased, objective, transparent, ecologically sound, and fiscally prudent.

The operational efficiency of waterway transport of forest chips on Finland's Lake Saimaa

Energy Technology Data Exchange (ETDEWEB)

Karttunen, K.; Ranta, T. [Lappeenranta Univ. of Technology, LUT Savo Sustainable Technologies, Mikkeli (Finland); Vaatainen, K.; Asikainen, A. [The Finnish Forest Research Inst., Joensuu (Finland)], E-mail: kalle.karttunen@lut.fi

2012-11-01

New and cost-efficient methods for use in supply chains for energy wood should be found, to reach the targets of the renewable energy utilisation set by the European Union. The long-distance waterway transportation of forest fuels should be thoroughly investigated, especially in areas where the transport distance is long and waterways could provide a feasible method of conveying forest fuel. In comparison to transport of forest chips by truck, barge-based waterway transport shows a competitive advantage due to the larger loads and higher bulk density of chips it allows. The cost-efficiency of waterway transportation operations related to forest chips in Finland's Lake Saimaa region was studied using practical demonstrations and discrete-event simulation. The varying demand for fuel wood in three separate bio-power plants on the Saimaa lakeside (near the cities of Varkaus, Mikkeli, and Savonlinna) was addressed in several barge transportation scenarios. Finally, the economy of barge transportation was compared to the economy of truck transportation as a function of transportation distance and in terms of the annual performance of the transportation methods examined. The waterway supply chain of forest chips was cost-competitive to road transport by truck after 100-150 km. According to the simulation study, the most economical waterway transport options were based on fixed barge system and shift-independent harbor logistics where loading and unloading of barges were carried-out with a wheeled loader and a belt conveyor. Total supply chain costs including the best waterway logistics from road side storage to power plant ranged from 10.75 euros to 11.64 euros/MWh in distances of 100-150 km by waterways. The energy-density of forest chips in the barge load was found to be, on average, 25% higher than that in truck hauling, because of the better compaction of chips. Waterway transport is a viable option for long-distance transportation of forest chips in Eastern

77 FR 69562 - Drawbridge Operation Regulation; Thea Foss Waterway Previously Known as City Waterway, Tacoma, WA

Science.gov (United States)

2012-11-20

... shape in the ``Actions'' column. If you submit your comments by mail or hand delivery, submit them in an... unique features of the Murray Morgan Bridge is its height above the waterway providing 60 feet of... bridge openings are for locally moored and operated recreational sailboats with mast heights over 60 feet...

78 FR 19277 - Navigation Safety Advisory Council; Meeting

Science.gov (United States)

2013-03-29

... regarding multiple risk assessments for U.S. ports and waterways. If you have been adversely affected by the... measures; marine information; diving safety; and aids to navigation systems. This notice corrects the...

76 FR 56638 - Safety Zone; Head of the Cuyahoga, Cuyahoga River, Cleveland, OH

Science.gov (United States)

2011-09-14

... likely combination of large numbers of recreational vessels, congested waterways, and alcohol use..., 81.40'50'' W (Marathon Bend) to a line drawn perpendicular to each river bank at 41.29'56'' N, 81.42... standards are technical standards (e.g., specifications of materials, performance, design, or operation...

Forstering cooperation in inland waterway networks : a gaming and simulation approach (chapter 23)

NARCIS (Netherlands)

Veenstra, A.W.; Meijeren, van J.; Harmsen, J.M.A.; Verbraeck, A.; Negenborn, R.R.; Ocampa-Martinez, C.

2015-01-01

Inland waterway transport is an important economic activity in the Netherlands and in Europe. Especially in the hinterland transport of containers, inland shipping is expected to form the backbone of a multimodal transport system. To support and strengthen the inland waterway industry in the

Relationships Between Perceived Coastal Waterway Condition and Social Aspects of Quality of Life

Directory of Open Access Journals (Sweden)

Melanie E. Cox

2006-06-01

Full Text Available Previous research has shown that the presence of natural areas is beneficial to human well-being. However, to date there have been few published studies of the effects that the condition of natural areas have on well-being. We hypothesize that coastal waterways that are perceived to be in better condition are visited more often by local residents, and as a result, residents will develop a stronger sense of place and stronger social relations with other residents, which will in turn lead to a higher quality of life. A survey was conducted to test this hypothesis in two coastal regions in Queensland, Australia. A weak relationship was found between perceived coastal quality and the number of recreational visits. In both study areas, frequency of visits to coastal waterways was significantly related to quality of life through an increased sense of place and social contacts. In the Douglas region, sense of place and social capital were closely related, and social capital was also related to quality of life. In both study areas, residents were most likely to visit waterways that were located in close proximity to their residence, suggesting that management of all waterways is essential for human well-being, and that benefits to humans will result from rehabilitating degraded waterways.

Ensuring the nuclear safety of VVER-440 reactor pressure vessels in Skoda, Concern Enterprise, Plzen

International Nuclear Information System (INIS)

Hrbek, Z.

1985-01-01

Various types of routine inspections are described of reactor pressure vessels with the aim of identifying residual lifetime and overall safety. The inspection programme includes: choice of systems and instruments, type of tests, test frequency, safety criteria, measures to be taken in case of unsatisfactory results, documentation. The criteria are given for periodical inspections and requirements listed for instruments and equipment. The main three groups of tests are: visual inspection and dimension tests, surface inspection and volumetric inspection. Briefly described is some of the equipment used. (M.D.)

33 CFR 165.720 - Safety/Security Zone: St. Johns River, Jacksonville, FL.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety/Security Zone: St. Johns... OF HOMELAND SECURITY (CONTINUED) PORTS AND WATERWAYS SAFETY REGULATED NAVIGATION AREAS AND LIMITED... § 165.720 Safety/Security Zone: St. Johns River, Jacksonville, FL. (a) Location. The water and the land...

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-01

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

Detection of multi-drug resistant Escherichia coli in the urban waterways of Milwaukee, WI

Directory of Open Access Journals (Sweden)

Anthony D. Kappell

2015-04-01

Full Text Available Urban waterways represent a natural reservoir of antibiotic resistance which may provide a source of transferable genetic elements to human commensal bacteria and pathogens. The objective of this study was to evaluate antibiotic resistance of Escherichia coli isolated from the urban waterways of Milwaukee, WI compared to those from Milwaukee sewage and a clinical setting in Milwaukee. Antibiotics covering 10 different families were utilized to determine the phenotypic antibiotic resistance for all 259 E. coli isolates. All obtained isolates were determined to be multi-drug resistant. The E. coli isolates were also screened for the presence of the genetic determinants of resistance including ermB (macrolide resistance, tet(M (tetracycline resistance, and β-lactamases (blaOXA, blaSHV, and blaPSE. E. coli from urban waterways showed a greater incidence of antibiotic resistance to 8 of 17 antibiotics tested compared to human derived sources. These E. coli isolates also demonstrated a greater incidence of resistance to higher numbers of antibiotics compared to the human derived isolates. The urban waterways demonstrated a greater abundance of isolates with co-occurrence of antibiotic resistance than human derived sources. When screened for 5 different antibiotic resistance genes conferring macrolide, tetracycline, and β-lactam resistance, clinical E. coli isolates were more likely to harbor ermB and blaOXA than isolates from urban waterway. These results indicate that Milwaukee’s urban waterways may select for a greater incidence of multiple antibiotic resistance organisms and likely harbor a different antibiotic resistance gene pool than clinical sources. The implications of this study are significant to understanding the presence of resistance in urban freshwater environments by supporting the idea that sediment from urban waterways serves as a reservoir of antibiotic resistance.

33 CFR 165.116 - Safety and Security Zones; Salem and Boston Harbors, Massachusetts.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety and Security Zones; Salem..., DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PORTS AND WATERWAYS SAFETY REGULATED NAVIGATION AREAS AND LIMITED... § 165.116 Safety and Security Zones; Salem and Boston Harbors, Massachusetts. (a) Location. The...

78 FR 59339 - Intracoastal Waterway Route “Magenta Line” on NOAA Nautical Charts

Science.gov (United States)

2013-09-26

... waterways have been based principally on surveys made from 60 to 80 years ago and, necessarily, are obsolete... found where the charted Intracoastal Waterway Route (``magenta line'') passes on the wrong side of aids... specifically seeks comments regarding: 1. How do you currently access the magenta line? On paper nautical...

Test Bed for Safety Assessment of New e-Navigation Systems

Directory of Open Access Journals (Sweden)

Axel Hahn

2014-12-01

Full Text Available New e-navigation strains require new technologies, new infrastructures and new organizational structures on bridge, on shore as well as in the cloud. Suitable engineering and safety/risk assessment methods facilitate these efforts. Understanding maritime transportation as a sociotechnical system allows the application of system-engineering methods. Formal, simulation based and in situ verification and validation of e-navigation technologies are important methods to obtain system safety and reliability. The modelling and simulation toolset HAGGIS provides methods for system specification and formal risk analysis. It provides a modelling framework for processes, fault trees and generic hazard specification and a physical world and maritime traffic simulation system. HAGGIS is accompanied by the physical test bed LABSKAUS which implements a physical test bed. The test bed provides reference ports and waterways in combination with an experimental Vessel Traffic Services (VTS system and a mobile integrated bridge: This enables in situ experiments for technological evaluation, testing, ground research and demonstration. This paper describes an integrated seamless approach for developing new e-navigation technologies starting with simulation based assessment and ending in physical real world demonstrations

Investigating driver willingness to drive through flooded waterways.

Science.gov (United States)

Pearson, Marti; Hamilton, Kyra

2014-11-01

Approximately 40% of all drowning deaths involve a motor vehicle. Regardless of its significance as a cause of flood-related mortality, there is continued prevalence of driving through flooded waterways in Australia and worldwide. We aimed to understand the motivational determinates of driving through flooded waterways in low and high-risk scenarios by utilizing an augmented theory of planned behaviour (TPB) with behavioural willingness as the outcome variable as well as the influence of additional predictors; namely perceived risk and past behaviour. Participants (n=174; Mage=27.43, SD=10.76) answered standard TPB-based questions in regards to attitudes, subjective norm, and perceived behavioural control (PBC), as well as additional variables of perceived risk (i.e., perceived susceptibility and perceived severity) and past behaviour. Support was found for the augmented TPB as attitude, subjective norm, and PBC predicted behavioural willingness. Support was also found for perceived severity in the high-risk but not the low-risk scenario. No support was found for perceived susceptibility. Past behaviour emerged as a significant predictor of willingness in the low and high-risk scenario. The findings provide support for an augmented TPB in understanding individuals' willingness to drive through flooded waterways, suggesting that a multi-strategy approach may be critical in attempts to reduce the incidence of such risky driving behaviour. Copyright © 2014 Elsevier Ltd. All rights reserved.

Autonomous Radiation Monitoring of Small Vessels

International Nuclear Information System (INIS)

Fabris, Lorenzo; Hornback, Donald Eric

2010-01-01

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

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.)

78 FR 42452 - Safety Zone; Kentucky Air National Guard Vessel for Parachute Rescue Jumpmaster Training, Lake...

Science.gov (United States)

2013-07-16

... only a few hours at a time during any 24 hour period. The majority of the training exercises will be... greatly reducing the likelihood of affecting transient recreational vessels. Additionally, the starting... Health Risks and Safety Risks. This rule is not an economically significant rule and does not create an...

77 FR 14959 - Special Local Regulation for Marine Event; Temporary Change of Dates for Recurring Marine Events...

Science.gov (United States)

2012-03-14

... to relatively small rowing vessels and high-powered racing vessels in restricted waterways make... proximity to relatively small rowing vessels and high-powered racing vessels in a restricted waterways make.... The category of water activities includes but is not limited to sail boat regattas, boat parades...

Passive safety features of low sodium void worth metal fueled cores in a bottom supported reactor vessel

International Nuclear Information System (INIS)

Chang, Y.I.; Marchaterre, J.F.; Wade, D.C.; Wigeland, R.A.; Kumaoka, Yoshio; Suzuki, Masao; Endo, Hiroshi; Nakagawa, Hiroshi

1991-01-01

A study has been performed on the passive safety features of low-sodium-void-worth metallic-fueled reactors with emphasis on using a bottom-supported reactor vessel design. The reactor core designs included self-sufficient types as well as actinide burners. The analyses covered the reactor response to the unprotected, i.e. unscrammed, transient overpower accident and the loss-of-flow accident. Results are given demonstrating the safety margins that were attained. 4 refs., 4 figs., 2 tabs

Quantitative Evaluation of Stereo Visual Odometry for Autonomous Vessel Localisation in Inland Waterway Sensing Applications

Directory of Open Access Journals (Sweden)

Thomas Kriechbaumer

2015-12-01

Full Text Available Autonomous survey vessels can increase the efficiency and availability of wide-area river environment surveying as a tool for environment protection and conservation. A key challenge is the accurate localisation of the vessel, where bank-side vegetation or urban settlement preclude the conventional use of line-of-sight global navigation satellite systems (GNSS. In this paper, we evaluate unaided visual odometry, via an on-board stereo camera rig attached to the survey vessel, as a novel, low-cost localisation strategy. Feature-based and appearance-based visual odometry algorithms are implemented on a six degrees of freedom platform operating under guided motion, but stochastic variation in yaw, pitch and roll. Evaluation is based on a 663 m-long trajectory (>15,000 image frames and statistical error analysis against ground truth position from a target tracking tachymeter integrating electronic distance and angular measurements. The position error of the feature-based technique (mean of ±0.067 m is three times smaller than that of the appearance-based algorithm. From multi-variable statistical regression, we are able to attribute this error to the depth of tracked features from the camera in the scene and variations in platform yaw. Our findings inform effective strategies to enhance stereo visual localisation for the specific application of river monitoring.

Comparison of Multi-Scale Digital Elevation Models for Defining Waterways and Catchments Over Large Areas

Science.gov (United States)

Harris, B.; McDougall, K.; Barry, M.

2012-07-01

Digital Elevation Models (DEMs) allow for the efficient and consistent creation of waterways and catchment boundaries over large areas. Studies of waterway delineation from DEMs are usually undertaken over small or single catchment areas due to the nature of the problems being investigated. Improvements in Geographic Information Systems (GIS) techniques, software, hardware and data allow for analysis of larger data sets and also facilitate a consistent tool for the creation and analysis of waterways over extensive areas. However, rarely are they developed over large regional areas because of the lack of available raw data sets and the amount of work required to create the underlying DEMs. This paper examines definition of waterways and catchments over an area of approximately 25,000 km2 to establish the optimal DEM scale required for waterway delineation over large regional projects. The comparative study analysed multi-scale DEMs over two test areas (Wivenhoe catchment, 543 km2 and a detailed 13 km2 within the Wivenhoe catchment) including various data types, scales, quality, and variable catchment input parameters. Historic and available DEM data was compared to high resolution Lidar based DEMs to assess variations in the formation of stream networks. The results identified that, particularly in areas of high elevation change, DEMs at 20 m cell size created from broad scale 1:25,000 data (combined with more detailed data or manual delineation in flat areas) are adequate for the creation of waterways and catchments at a regional scale.

77 FR 55439 - Vessel Traffic Service Updates, Including Establishment of Vessel Traffic Service Requirements...

Science.gov (United States)

2012-09-10

... foreign-flag shipping industry were tasked to identify the information needs of waterway users to help... stakeholders to help identify needed VTM improvements and to determine candidate VTS waterways. PAWSA provides... are given appropriate attention as risk-mitigation measures are selected. The Coast Guard has...

Model-Based Design and Formal Verification Processes for Automated Waterway System Operations

Directory of Open Access Journals (Sweden)

Leonard Petnga

2016-06-01

Full Text Available Waterway and canal systems are particularly cost effective in the transport of bulk and containerized goods to support global trade. Yet, despite these benefits, they are among the most under-appreciated forms of transportation engineering systems. Looking ahead, the long-term view is not rosy. Failures, delays, incidents and accidents in aging waterway systems are doing little to attract the technical and economic assistance required for modernization and sustainability. In a step toward overcoming these challenges, this paper argues that programs for waterway and canal modernization and sustainability can benefit significantly from system thinking, supported by systems engineering techniques. We propose a multi-level multi-stage methodology for the model-based design, simulation and formal verification of automated waterway system operations. At the front-end of development, semi-formal modeling techniques are employed for the representation of project goals and scenarios, requirements and high-level models of behavior and structure. To assure the accuracy of engineering predictions and the correctness of operations, formal modeling techniques are used for the performance assessment and the formal verification of the correctness of functionality. The essential features of this methodology are highlighted in a case study examination of ship and lock-system behaviors in a two-stage lock system.

33 CFR 165.1318 - Security and Safety Zone Regulations, Large Passenger Vessel Protection, Portland, OR Captain of...

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Security and Safety Zone Regulations, Large Passenger Vessel Protection, Portland, OR Captain of the Port Zone 165.1318 Section 165.1318 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PORTS AND...

Aseismic safety analysis of a prestressed concrete containment vessel for CPR1000 nuclear power plant

Science.gov (United States)

Yi, Ping; Wang, Qingkang; Kong, Xianjing

2017-01-01

The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete containment vessel (PCCV) model with a semi-infinite elastic foundation and practical arrangement of tendons has been established to analyze the aseismic ability of the CPR1000 PCCV structure under seismic loads and internal pressure. A method to model the prestressing tendon and its interaction with concrete was proposed and the axial force of the prestressing tendons showed that the simulation was reasonable and accurate. The numerical results show that for the concrete structure, the location of the cylinder wall bottom around the equipment hatch and near the ring beam are critical locations with large principal stress. The concrete cracks occurred at the bottom of the PCCV cylinder wall under the peak earthquake motion of 0.50 g, however the PCCV was still basically in an elastic state. Furthermore, the concrete cracks occurred around the equipment hatch under the design internal pressure of 0.4MPa, but the steel liner was still in the elastic stage and its leak-proof function soundness was verified. The results provide the basis for analysis and design of containment vessels.

COMPARISON OF MULTI-SCALE DIGITAL ELEVATION MODELS FOR DEFINING WATERWAYS AND CATCHMENTS OVER LARGE AREAS

Directory of Open Access Journals (Sweden)

B. Harris

2012-07-01

Full Text Available Digital Elevation Models (DEMs allow for the efficient and consistent creation of waterways and catchment boundaries over large areas. Studies of waterway delineation from DEMs are usually undertaken over small or single catchment areas due to the nature of the problems being investigated. Improvements in Geographic Information Systems (GIS techniques, software, hardware and data allow for analysis of larger data sets and also facilitate a consistent tool for the creation and analysis of waterways over extensive areas. However, rarely are they developed over large regional areas because of the lack of available raw data sets and the amount of work required to create the underlying DEMs. This paper examines definition of waterways and catchments over an area of approximately 25,000 km2 to establish the optimal DEM scale required for waterway delineation over large regional projects. The comparative study analysed multi-scale DEMs over two test areas (Wivenhoe catchment, 543 km2 and a detailed 13 km2 within the Wivenhoe catchment including various data types, scales, quality, and variable catchment input parameters. Historic and available DEM data was compared to high resolution Lidar based DEMs to assess variations in the formation of stream networks. The results identified that, particularly in areas of high elevation change, DEMs at 20 m cell size created from broad scale 1:25,000 data (combined with more detailed data or manual delineation in flat areas are adequate for the creation of waterways and catchments at a regional scale.

75 FR 5511 - Safety Zone; AICW Closure Safety Zone for Ben Sawyer Bridge Replacement Project, Sullivan's...

Science.gov (United States)

2010-02-03

..., including potential falling debris and the use of heavy equipment and machinery in the waterway, could lead... operation presents a potential hazard to mariners from falling debris and the use of heavy equipment and machinery. To provide for the safety of the public, the Coast Guard will temporarily restrict access to this...

76 FR 58112 - Safety Zone; M/V DAVY CROCKETT, Columbia River

Science.gov (United States)

2011-09-20

... Silvestre Suga, Waterways Management Division, Marine Safety Unit Portland, Coast Guard; telephone 503-240- 9319, e-mail Silvestre[email protected] . If you have questions on viewing the docket, call Renee V...

33 CFR 165.904 - Lake Michigan at Chicago Harbor & Burnham Park Harbor-Safety and Security Zone.

Science.gov (United States)

2010-07-01

... & Burnham Park Harbor-Safety and Security Zone. 165.904 Section 165.904 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PORTS AND WATERWAYS SAFETY REGULATED NAVIGATION... Guard District § 165.904 Lake Michigan at Chicago Harbor & Burnham Park Harbor—Safety and Security Zone...

33 CFR 165.731 - Safety/Security Zone: Cumberland Sound, Georgia and St. Marys River Entrance Channel.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety/Security Zone: Cumberland... Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PORTS AND WATERWAYS SAFETY REGULATED... Seventh Coast Guard District § 165.731 Safety/Security Zone: Cumberland Sound, Georgia and St. Marys River...

78 FR 44011 - Safety Zone; Bullhead City Regatta; Bullhead City, AZ

Science.gov (United States)

2013-07-23

... for this large waterway user marine event. C. Discussion of Comments, Changes and the Final Rule The... excessive trash resulting from the event. The safety zone is being established this year as in past years to... safety measures have been taken and fees associated with the event are used for trash clean up...

TPE upgrade for enhancing operational safety and improving in-vessel tritium inventory assessment in fusion nuclear environment

Energy Technology Data Exchange (ETDEWEB)

Shimada, M., E-mail: Masashi.Shimada@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Taylor, C.N.; Moore-McAteer, L.; Pawelko, R.J. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Kolasinski, R.D.; Buchenauer, D.A. [Sandia National Laboratories, Hydrogen and Materials Science Department, Livermore, CA 94550 (United States); Cadwallader, L.C.; Merrill, B.J. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

2016-11-01

The Tritium Plasma Experiment (TPE) is a unique high-flux linear plasma device that can handle beryllium, tritium, and neutron-irradiated plasma facing materials, and is the only existing device dedicated to evaluate in-vessel tritium inventory in the nuclear environment for fusion safety. The electrical upgrade were recently carried out to enhance operational safety and to improve plasma performance. New DC power supplies and a new control center enable remote plasma operations from outside of the contamination area for tritium, minimizing the possible exposure risk with tritium and beryllium and eliminating heat stress issue. In November 2015, the TPE successfully achieved first deuterium plasma via remote operation after a significant three-year upgrade. Simple linear scaling estimate showed that the TPE is expected to achieve Γ{sub i}{sup max} of >1.0 × 10{sup 23} m{sup −2} s{sup −1} and q{sub heat} of >1 MW m{sup −2} with new power supplies. This upgrade not only improves operational safety of the worker, but also enhances plasma performance to better simulate extreme plasma-material conditions expected in ITER, FNSF, and DEMO for improving in-vessel tritium inventory assessment in fusion nuclear environment.

Cooling system for the connecting rings of a fast neutron reactor vessel

International Nuclear Information System (INIS)

Martin, J.-P.; Malaval, Claude

1974-01-01

A description is given of a cooling system for the vessel connecting rings of a fast neutron nuclear reactor, particularly of a main vessel containing the core of the reactor and a volume of liquid metal coolant at high temperature and a safety vessel around the main vessel, both vessels being suspended to a rigid upper slab kept at a lower temperature. It is mounted in the annular space between the two vessels and includes a neutral gas circuit set up between the wall of the main vessel to be cooled and that of the safety vessel itself cooled from outer. The neutral gas system comprises a plurality of ventilators fitted in holes made through the thickness of the upper slab and opening on to the space between the two vessels. It also includes two envelopes lining the walls of these vessels, establishing with them small section channels for the circulation of the neutral gas cooled against the safety vessel and heated against the main vessel [fr

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

76 FR 73511 - Safety Zone; M/V DAVY CROCKETT, Columbia River

Science.gov (United States)

2011-11-29

... INFORMATION CONTACT: If you have questions on this temporary rule, call or email BM1 Silvestre Suga, Waterways Management Division, Marine Safety Unit Portland, Coast Guard; telephone (503) 240-9319, email Silvestre.G...

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.

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

International Nuclear Information System (INIS)

1975-12-01

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

75 FR 23793 - Houston/Galveston Navigation Safety Advisory Committee

Science.gov (United States)

2010-05-04

... DEPARTMENT OF HOMELAND SECURITY Coast Guard [Docket No. USCG-2010-0032] Houston/Galveston... Houston/Galveston Navigation Safety Advisory Committee (``HOGANSAC'' or ``the Committee'') and its working groups will meet in Houston, Texas to discuss waterway improvements, aids to navigation, area projects...

In service inspection of SUPERPHENIX 1 vessels: MIR

International Nuclear Information System (INIS)

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

1985-01-01

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

76 FR 49664 - Safety Zone; M/V DAVY CROCKETT, Columbia River

Science.gov (United States)

2011-08-11

... questions on this temporary rule, call or e-mail BM1 Silvestre Suga, Waterways Management Division, Marine Safety Unit Portland, Coast Guard; telephone 503-240- 9319, e-mail Silvestre[email protected] . If you...

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.)

Socio-cultural impacts of construction and regulation of the waterway of Oulujoki river

International Nuclear Information System (INIS)

Ruotsala, H.

1992-01-01

The purpose of study was to clarify the socio-cultural impacts of the regulation and construction of Oulujoki river on the everyday life of people. Changes in work and livelihood, living and living satisfaction, and leisure time and recreation are considered as variables. Quantitative and qualitative methods were combined in the study. The principal study material is composed of answers to a questionnaire and of 109 thematic interviews on the topic. The hold of the study is phenomenological-hermeneutical, with the intention to interpret the matter from the viewpoint of the attitudes and requirements of the waterway users. The theoretical frame of reference was culture-ecological point of view and adaptive process. The emphasis of the study is on sociocultural adaptation. The attitude towards waterway construction and regulation was dependent on the background of the person interviewed. The attitude was considered e.g. by various interest groups and professional groups, such as farmers, inhabitants on the shore, recreation users, fishermen, municipal elected officials. Other important variables were age, family stage and sex. Big local differences in the attitudes could also be found. While prevailing values in society and peoples' modes of living are changing, also the attitude towards waterway and changes caused by waterway construction has changed in the region studied. The impacts on the sources of livelihood are minor for the moment, the impacts on living satisfaction and recreations are significant. While leisure time is increasing the recreation value of waterway increases, too. At the construction stage of the river, the economical and productional advantages took priority, and then no attention was paid to recreational use

Application of GIS tools in determining the navigability of waterways

Science.gov (United States)

Nadolny, Grzegorz; Rabant, Hubert; Szatten, Dawid

2017-11-01

This article presents the results of a research conducted on Lower Noteć river for the application of geographic information system (GIS) tools. The study consisted of longitudinal profile soundings of navigable route combined with statistical analysis of water levels. GIS software - ArcMap v. 10.0 was used to perform analysis of changes in waterway depth depending on hydrological conditions. A mileage of waterway sections was specified depending on whether they met or did not meet classification requirements in accordance with Polish law. The application of spatial data of Lower Noteć river developed for the purpose of the article is presented. Conducted analyses and obtained results demonstrate the importance of GIS tools in inland navigation studies.

77 FR 42649 - Safety Zone: Sea World San Diego Fireworks, Mission Bay; San Diego, CA

Science.gov (United States)

2012-07-20

... 1625-AA00 Safety Zone: Sea World San Diego Fireworks, Mission Bay; San Diego, CA AGENCY: Coast Guard... authorized by the Captain of the Port, or his designated representative. DATES: This rule is effective from 8... to ensure the public's safety. B. Basis and Purpose The Ports and Waterways Safety Act gives the...

Evaluation of the Structural Safety of a Vessel with Different Material(Cr-13)-Supplemented Screw Thread

Energy Technology Data Exchange (ETDEWEB)

Choi, Yong Hoon; Bae, Jun Ho; Kim, Chul [Pusan National University, Busan (Korea, Republic of)

2015-04-15

The dome and neck part of a vessel is generally formed by a hot spinning process with a seamless tube. However, as studies on and design data from the hot spinning process are insufficient, this process has been performed based on trial and error and the experiences of field engineers. Changes in the inner diameter from the bottom to the top of the neck have occurred mainly because of the characteristics of the hot spinning process due to the high-speed rotation of the rollers. In this study, a theoretical and finite element analysis of the vessel is conducted with different material(Cr-13)-supplemented screw threads for tapping and to reduce shape errors. Based on the results, the structural safety under the operating conditions is evaluated.

Proximal Occlusion of Medium-Sized Vessels with the Penumbra Occlusion Device: A Study of Safety and Efficacy

Energy Technology Data Exchange (ETDEWEB)

Jambon, E.; Petitpierre, F. [Pellegrin Hospital, Department of Radiology (France); Brizzi, V.; Dubuisson, V. [Pellegrin Hospital, Department of Surgery (France); Bras, Y. Le; Grenier, N.; Cornelis, F., E-mail: cornelisfrancois@gmail.com [Pellegrin Hospital, Department of Radiology (France)

2017-02-15

PurposeTo retrospectively investigate the safety and efficacy of hybrid proximal coiling of various medium-sized vessels (4 to 8 mm) using the Penumbra Occlusion Device (POD).Materials and MethodsFrom October 2014 to February 2016, 37 proximal embolizations were performed with PODs in 36 patients (mean age: 50.8, range: 10–86; 29 male, 7 female). Vessel occlusions were achieved under fluoroscopic guidance using a 2.7 French microcatheter. Among the 36 vessels targeted, 16 were splenic arteries, 11 renal arteries, 4 mesenteric arteries, 3 arteriovenous fistulae, 1 iliac artery, and 1 gonadal vein. Intermittent follow-up angiography was performed to assess the flow for final occlusion. Outcomes and complications were assessed by clinical and/or imaging follow-up.ResultsTo produce proximal occlusion of the intended vessels, the POD was used alone in 19 embolizations (51.4 %). In 12 procedures (32.4 %), POD was used as a coil constrainer to secure the coil construct. In 6 procedures (16.2 %), additional embolic devices were used to achieve vessel occlusion after initial POD deployment. After a mean follow-up of 3.2 months, no POD migration was observed but two complications occurred (5.4 %): one post embolic syndrome and one extensive infarction with splenic abscess.ConclusionThe POD system allows safe and effective proximal embolization of medium-sized vessels in a variety of clinical settings.

Assessment of the Zaporizhya NPP unit 1 reactor pressure vessel safety

International Nuclear Information System (INIS)

Podkopaev, V.; Popov, V.; Zaritsky, N.

1997-01-01

This emergency situation had occurred at the ZNPP unit 1 while its being under ''hot shutdown'' in natural coolant circulation mode. The main difference between emergency situation and mode with improper setting of PPPD described in the ''Technical Safety Substantiation (TSS) is that this mode is being considered in the TSS under rated power of reactor with main circulation pumps (MCP) under operation. This difference is a substantial one. For this reason a necessity appeared to asses an integrity of referred reactor pressure vessel (RPV) under given emergency situation to judge whether results obtained meet the ND requirements (safety assessment). Under operation such RPV elements are being mostly affected as upper cooling, lower cowling, weld No. 3 weld No. 4 situated in front of core. These elements materials ageing process is the most intense one. Thus, this work was aimed at investigation of structure material behavior and RPV integrity assessment under thermal shock conditions while PPPD improper setting. At that time the most attention was drawn to above mentioned upper and lower cowlings along with welds No. 3 and 4. 5 refs, figs, 10 tabs

Assessment of the Zaporizhya NPP unit 1 reactor pressure vessel safety

Energy Technology Data Exchange (ETDEWEB)

Podkopaev, V; Popov, V; Zaritsky, N [State Scientific and Technical Centre on Nuclear and Radiation Safety (SSTC NRS), Kiev (Ukraine)

1997-09-01

This emergency situation had occurred at the ZNPP unit 1 while its being under ``hot shutdown`` in natural coolant circulation mode. The main difference between emergency situation and mode with improper setting of PPPD described in the ``Technical Safety Substantiation (TSS) is that this mode is being considered in the TSS under rated power of reactor with main circulation pumps (MCP) under operation. This difference is a substantial one. For this reason a necessity appeared to asses an integrity of referred reactor pressure vessel (RPV) under given emergency situation to judge whether results obtained meet the ND requirements (safety assessment). Under operation such RPV elements are being mostly affected as upper cooling, lower cowling, weld No. 3 weld No. 4 situated in front of core. These elements materials ageing process is the most intense one. Thus, this work was aimed at investigation of structure material behavior and RPV integrity assessment under thermal shock conditions while PPPD improper setting. At that time the most attention was drawn to above mentioned upper and lower cowlings along with welds No. 3 and 4. 5 refs, figs, 10 tabs.

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-02-01

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

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

International Nuclear Information System (INIS)

Monteleone, S.

1997-02-01

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

33 CFR 165.502 - Safety and Security Zone; Cove Point Liquefied Natural Gas Terminal, Chesapeake Bay, Maryland.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety and Security Zone; Cove... Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PORTS AND WATERWAYS SAFETY... Areas Fifth Coast Guard District § 165.502 Safety and Security Zone; Cove Point Liquefied Natural Gas...

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

International Nuclear Information System (INIS)

1982-10-01

The central problem in the unresolved safety issue A-11, Reactor Vessel Materials Toughness, was to provide guidance in performing analyses required by 10 CFR Part 50, Appendix G, Section V.C. for reactor pressure vessels (RPVs) which fail to meet the toughness requirement during service life as a result of neutron radiation embrittlement. Although the methods of linear-elastic fracture mechanics (LEFM) were adequate for low-temperature RPV problems, they were inapplicable under operating conditions because vessel steels, even those which exhibit less than 50 ft-lb of C/sub v/ energy, were relatively tough at temperatures where the impact energy reached its upper shelf values. 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 had been developed under earlier NRC sponsorship, relationships were obtained which provided approximate solutions to the problem of RPV fracture with assumed beltline region flaws. The first paper of this report is a summary of the problem, the solutions, and the results of verification analyses. The details are provided in a series of appendices in Volumes I and II

Hydrology of and Current Monitoring Issues for the Chicago Area Waterway System, Northeastern Illinois

Science.gov (United States)

Duncker, James J.; Johnson, Kevin K.

2015-10-28

The Chicago Area Waterway System (CAWS) consists of a combination of natural and manmade channels that form an interconnected navigable waterway of approximately 90-plus miles in the metropolitan Chicago area of northeastern Illinois. The CAWS serves the area as the primary drainage feature, a waterway transportation corridor, and recreational waterbody. The CAWS was constructed by the Metropolitan Water Reclamation District of Greater Chicago (MWRDGC). Completion of the Chicago Sanitary and Ship Canal (initial portion of the CAWS) in 1900 breached a low drainage divide and resulted in a diversion of water from the Lake Michigan Basin. A U.S. Supreme Court decree (Consent Decree 388 U.S. 426 [1967] Modified 449 U.S. 48 [1980]) limits the annual diversion from Lake Michigan. While the State of Illinois is responsible for the diversion, the MWRDGC regulates and maintains water level and water quality within the CAWS by using several waterway control structures. The operation and control of water levels in the CAWS results in a very complex hydraulic setting characterized by highly unsteady flows. The complexity leads to unique gaging requirements and monitoring issues. This report provides a general discussion of the complex hydraulic setting within the CAWS and quantifies this information with examples of data collected at a range of flow conditions from U.S. Geological Survey streamflow gaging stations and other locations within the CAWS. Monitoring to address longstanding issues of waterway operation, as well as current (2014) emerging issues such as wastewater disinfection and the threat from aquatic invasive species, is included in the discussion.

33 CFR 88.11 - Law enforcement vessels.

Science.gov (United States)

2010-07-01

... NAVIGATION RULES ANNEX V: PILOT RULES § 88.11 Law enforcement vessels. (a) Law enforcement vessels may display a flashing blue light when engaged in direct law enforcement or public safety activities. This... lights. (b) The blue light described in this section may be displayed by law enforcement vessels of the...

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

2011 Vessel Density

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

2013 Vessel Density

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

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

33 CFR 165.T11-304 - Safety zone; Sea World Summer Nights Fireworks; Mission Bay, San Diego, California.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety zone; Sea World Summer... Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PORTS AND WATERWAYS SAFETY... Areas Eleventh Coast Guard District § 165.T11-304 Safety zone; Sea World Summer Nights Fireworks...

33 CFR 165.1182 - Safety/Security Zone: San Francisco Bay, San Pablo Bay, Carquinez Strait, and Suisun Bay, CA.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety/Security Zone: San... Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PORTS AND WATERWAYS SAFETY... Areas Eleventh Coast Guard District § 165.1182 Safety/Security Zone: San Francisco Bay, San Pablo Bay...

76 FR 2916 - Houston/Galveston Navigation Safety Advisory Committee; Meetings

Science.gov (United States)

2011-01-18

... DEPARTMENT OF HOMELAND SECURITY Coast Guard [USCG-2010-1116] Houston/Galveston Navigation Safety Advisory Committee; Meetings AGENCY: Coast Guard, DHS. ACTION: Notice of Meetings. SUMMARY: The Houston... will meet in Texas City, Texas and Houston, Texas to discuss waterway improvements, aids to navigation...

75 FR 5769 - Inland Waterways Users Board; Request for Nominations

Science.gov (United States)

2010-02-04

... ``various'' regions. For the purpose of selecting Board members, the waterways subjected to fuel taxes and... Products; (2) Coal and Coke; (3) Petroleum, Crude and Products; (4) Minerals, Ores, and Primary Metals and...

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

International Nuclear Information System (INIS)

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

2012-05-01

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

33 CFR 165.169 - Safety and Security Zones: New York Marine Inspection Zone and Captain of the Port Zone.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety and Security Zones: New... Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PORTS AND WATERWAYS SAFETY... Areas First Coast Guard District § 165.169 Safety and Security Zones: New York Marine Inspection Zone...

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.)

33 CFR 165.154 - Safety and Security Zones: Long Island Sound Marine Inspection Zone and Captain of the Port Zone.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety and Security Zones: Long... Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PORTS AND WATERWAYS SAFETY... Areas First Coast Guard District § 165.154 Safety and Security Zones: Long Island Sound Marine...

77 FR 47826 - Inland Waterways Users Board; Request for Nominations

Science.gov (United States)

2012-08-10

... the Board, the waterways subjected to fuel taxes and described in Public Law 95-502, as amended, have... United States. These categories are (1) Farm and Food Products; (2) Coal and Coke; (3) Petroleum, Crude...

Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessel internals

International Nuclear Information System (INIS)

2005-10-01

. The guidance reports are directed at technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant components addressed in the reports. The report addresses the reactor pressure vessel internals in BWRs. Maintaining the structural integrity of these reactor pressure vessel internals throughout NPP service life, in spite of several ageing mechanisms, is essential for plant safety

Nuclear power plant pressure vessels. Control of piping

International Nuclear Information System (INIS)

2000-01-01

The guide presents requirements for the pipework of nuclear facilities in Finland. According to the section 117 of the Finnish Nuclear Energy Degree (161/88), the Radiation and Nuclear Safety Authority of Finland (STUK) controls the pressure vessels of nuclear facilities in accordance with the Nuclear Energy Act (990/87) and, to the extent applicable in accordance with the Act of Pressure Vessels (98/73) and the rules and regulations issued by the virtue of these. In addition STUK is an inspecting authority of pressure vessels of nuclear facilities in accordance with the Pressure Vessel Degree (549/1973). According to the section of the Pressure Vessel Degree, a pressure vessel is a steam boiler, pressure container, pipework of other such appliance in which the pressure is above or may come to exceed the atmospheric pressure. Guide YVL 3.0 describes in general terms how STUK controls pressure vessels. STUK controls Safety Class 1, 2 and 3 piping as well as Class EYT (non-nuclear) and their support structures in accordance with this guide and applies the provisions of the Decision of the Ministry of Trade and Industry on piping (71/1975) issued by virtue of the Pressure Vessel Decree

PROPOSED GUIDELINES ON DEVELOPING THE OPTIMISATION MODEL FOR PASSAGE PLANNING IN INLAND WATERWAYS NAVIGATION

OpenAIRE

Ivančić, Paško; Kasum, Joško; Pavić, Ivica

2013-01-01

Road transport networks are overloaded by the increasing number of vehicles [1]. This has resulted in the redirection and the increasing intensity of navigation on rivers, canals and lakes. Transportation by inland waterways of the European Union (EU) represents a reliable, economical and environmentally friendly mode of transport. The latter has been recognized as a key form of transport in the European intermodal transport system. Inland waterways as part of the transport system play an imp...

The Analysis of the Causes of Emergencies on the Vessels

Directory of Open Access Journals (Sweden)

Alicja Mrozowska

2017-12-01

Full Text Available The article discusses the results of research conducted on the vessels, covering a wide spectrum of issues relating to the exploitation of vessels of various flags, as well as operating security and safety systems on board. The main aim of the study was to collect numbers of data directly from the crew, for examples: indicate by the crew marine areas with the greatest probability of occurrence of casualties and incidents, trying to the definition the causes of their occurrence, prevention actions used on board and analyses operating safety systems used on the various type of vessels. The analysis of research became the basis to identify strengths and weaknesses areas of the vessel operation. The author proposes a solution to be implemented on board and emphasizes meaning of safety management system.

75 FR 56015 - Vessel Inspection Alternatives

Science.gov (United States)

2010-09-15

... DEPARTMENT OF HOMELAND SECURITY Coast Guard 46 CFR Part 8 Vessel Inspection Alternatives CFR... Certificate; (ii) International Tonnage Certificate; (iii) Cargo Ship Safety Construction Certificate; (iv) Cargo Ship Safety Equipment Certificate; and (v) International Oil Pollution Prevention Certificate; and...

Analysis of stress in reactor core vessel under effect of pressure lose shock wave

International Nuclear Information System (INIS)

Li Yong; Liu Baoting

2001-01-01

High Temperature gas cooled Reactor (HTR-10) is a modular High Temperature gas cooled Reactor of the new generation. In order to analyze the safety characteristics of its core vessel in case of large rupture accident, the transient performance of its core vessel under the effect of pressure lose shock wave is studied, and the transient pressure difference between the two sides of the core vessel and the transient stresses in the core vessel is presented in this paper, these results can be used in the safety analysis and safety design of the core vessel of HTR-10. (author)

Efficiency and safety of bipolar vessel and tissue sealing in visceral surgery.

Science.gov (United States)

Overhaus, Marcus; Schaefer, Nico; Walgenbach, Klaus; Hirner, Andreas; Szyrach, Mara Natascha; Tolba, René Hany

2012-11-01

The aim of this study was to analyze the efficiency and safety of the bipolar tissue/vessel sealing and cutting device EnSeal(™) in comparison to the conventional clamp and ligation technique in visceral surgery. In an acute animal model, a part of the small bowel, a part of the colon and the kidneys were resected either with the conventional clamp and ligation technique or with EnSeal(™). Operation time, blood loss and blood parameters as well as the lateral thermal spread were evaluated. Small bowel, colon and kidney resection time with the EnSeal(™) device was shorter compared to the conventional clamp and ligation technique (small bowel: EnSeal(™): 4.7 ± 1.0 min vs. con: 35.1 ± 2.3 min; colon: EnSeal(™): 7.0 ± 1.4 min vs. con: 16.3 ± 1.5 min, kidney: EnSeal(™): 5.7 ± 1.3 min vs. con: 16.7 ± 3.7 min, p surgery with EnSeal(™) can be performed more efficiently in a shorter time, with significantly less blood loss, minimal thermal damage and without changes of blood parameters, indicating biological safety and integrity.

Assessment and management of ageing of major nuclear power plant components important to safety: PWR vessel internals: 2007 update

International Nuclear Information System (INIS)

2007-06-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that effective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) is one of the most important issues for plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wearout of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. IAEA-TECDOC-1119 documents ageing assessment and management practices for PWR Reactor Vessel Internals (RVIs) that were current at the time of its finalization in 1997-1998. Safety significant operating events have occurred since the finalization of the TECDOC, e.g. irradiation assisted stress corrosion cracking (IASCC) of baffle-former bolts, which threatened the integrity of the vessel internals. In addition, concern of fretting wear of control rod guide tubes has been raised in Japan. These events led to new ageing management actions by both NPP operators and regulators. Therefore it was recognized that IAEA-TECDOC-1119 should be updated by incorporating those new events and their countermeasures. The objective of this report is to update relevant sections of the existing IAEA-TECDOC- 1119 in order to provide current ageing management guidance for PWR RVIs to all involved in the operation and regulation of PWRs and thus to help ensure PWR safety in IAEA Member States throughout their entire service life

Influence of shock waves as a result of assumed vessel failure on parts of the plant relevant to safety

International Nuclear Information System (INIS)

Danisch, R.; Graubner, U.

1981-01-01

The shock wave induced rupture is of subordinate importance for the laying out of the parts of the plant relevant to safety. It is covered by the precautions for maximum potential earthquakes, aircraft crashes and chemical explosions. The failure of vessels in the power house (WAZUe, SPWB) as the result of a maximum potential earthquake is extremely improbable. If a combination of the stresses resulting from maximum potential earthquakes with the hypothetical stresses resulting from vessel failure is undertaken, it can be seen that the total stresses are only increased by a minimal amount, due to the quadratic averaging of less than 3%. (orig./DG) [de

Conditions Of Safe Ship Operations And Sea Waterway Parameters Based On The Reconstruction Of Świnoujście-Szczecin Fairway

Directory of Open Access Journals (Sweden)

Gucma Stanisław

2015-09-01

Full Text Available This paper describes the state vector of safe ship operation on sea waterways. The author determines relationships between the state vector of safe ship operation and parameters of sea waterway system elements. These include three subsystems: area, navigation and traffic control. These relationships have been exemplified with the 68-kilometre Świnoujście-Szczecin fairway, presently under reconstruction. A systematic approach to sea waterway design is globally unique.

33 CFR 165.T09-0073 - Safety and Security Zones; Tall Ships Challenge 2010; Great Lakes; Cleveland, OH; Bay City, MI...

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety and Security Zones; Tall... GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PORTS AND WATERWAYS SAFETY REGULATED NAVIGATION AREAS... Guard District § 165.T09-0073 Safety and Security Zones; Tall Ships Challenge 2010; Great Lakes...

2013 Cargo Vessel Density

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

2013 Fishing Vessel Density

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

2013 Tanker Vessel Density

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

2011 Fishing Vessel Density

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

Design of Eco Friendly Shallow Draft Fishing Vessel

Directory of Open Access Journals (Sweden)

Sunardi Sunardi

2016-04-01

Full Text Available One of the main problem of inland waterways fisheries is the transportation of fish from ponds to fish market during low tide trough inland waterways with 0.6m water depth.The boat is experiences grounding due to water depth of the river is not sufficient for the fishing boat to carry fish at it’s maximum2 tones capacity or experience dead freight . This condition forces fisherman to wait until the high tide from the sea, this delay causes the quality of the fish is decreasing.Besides the problem dead freight  problem the existing vessel is causes environmental problem such as erosion of the river bank due to wake wash. The other important issue is the increases of fuel price and it’s scarcity.  This paper presents the results of comparison of existing monohull fishing boat and two other alternativecatamaran designs. The catamaran design alternatives are is ordinary catamaran and flat side catamaran.  Both of the catamaran fishing boat design shows that the catamaran boat with 0.5m draft is able to carry more than 2 tonnes payload during low tide water depth.  The CFD simulation results shows that flat side catamaran resistance is more than 17.7% lower compared to ordinary catamaran and 44% lower compared to monohull. It means that the consumption of flat side catamaran is lowest compared to two other type of hull design. The flat side catamaran also produces lowest wake wash compared to o two other design. The low wake wash means more friendly to environment.

77 FR 31493 - Safety Zones; Fourth of July Fireworks Displays Within the Captain of the Port Charleston Zone, SC

Science.gov (United States)

2012-05-29

... temporary safety zones during Fourth of July Fireworks Displays on certain navigable waterways in Hilton... below. 1. Hilton Head Island, South Carolina. All waters within a 500 yard radius around the barge from... each safety zone. All coordinates are North American Datum 1983. [[Page 31496

In-service inspection robot for PFBR main vessel- concept

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

In-service inspection robot for PFBR main vessel- concept

International Nuclear Information System (INIS)

Rajendran, S.; Ramakumar, M.S.

1994-01-01

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

Application of probabilistic fracture mechanics to reactor pressure vessel safety assessment

International Nuclear Information System (INIS)

Venturini, V.; Pitner, P.

1995-06-01

Among all the components of a PWR (Pressurized Water Reactor) nuclear power plant, the reactor vessel is of major importance for safety. The integrity of this structure must be guaranteed in all circumstances, even in the case of the most severe accidents, and its mechanical state can be decisive for the lifetime of the plant. The brittle rupture would be the most important of all potential hazards if the irradiation effects were not consistent with predictions. The interest of having a reliable and precise method of evaluating the available safety margins and the integrity of this component led Electricite de France (EDF) to carry out a probabilistic fracture mechanics analysis. The probabilistic model developed by integration of the uncertainties in the usual fracture mechanics equations is presented. A special focus is made on the problem of coupling thermo-mechanical finite element calculations and reliability analysis. The use of a finite element code can be associated with prohibitive computation times when it is invoked numerous times during simulations sequences or complex iterative procedures. The response surface method is used. It provides an approximation of the response from a reduced number of original data. The global approach is illustrated on an example corresponding to a specific accidental transient. A validation of the obtained results is also carried out through the comparison with an equivalent model without coupling. (author)

MIR: an in-service inspection device for Superphenix 1 vessels

International Nuclear Information System (INIS)

Asty, M.; Ceccato, S.; Lerat, B.; Viard, J.

1986-06-01

The main and safety vessels of SUPERPHENIX 1 were designed to allow in-service inspections. The remote controlled inspection device MIR was developed for this purpose. It allows both visual and ultrasonic examinations to be performed. Basically, MIR consists of a tetrahedral structure provided with four steering and traction wheels, two for each vessel. A computer assisted control system enables it to be driven to any position on either the main or safety vessels. Operating conditions are briefly reviewed and the main features of MIR presented

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

Integrity of Magnox reactor steel pressure vessels

International Nuclear Information System (INIS)

Flewitt, P.E.J.; Williams, G.H.; Wright, M.B.

1992-01-01

The background to the safety assessment of the steel reactor pressure vessels for Magnox power stations is reviewed. The evolved philosophy adopted for the 1991 safety cases prepared for the continued operation of four Magnox power stations operated by Nuclear Electric plc is described, together with different aspects of the multi-legged integrity argument. The main revisions to the materials mechanical property data are addressed together with the assessment methodology adopted and their implications for the overall integrity argument formulated for the continued safe operation of these reactor pressure vessels. (author)

System for cooling the upper wall of a nuclear reactor vessel

International Nuclear Information System (INIS)

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

1974-01-01

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

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...

U.S. Navigated Waterway Mile Marker Locations from USCOE source data, Geographic NAD83, LOSCO (2000) [river_mile_markers_USACE_1994

Data.gov (United States)

Louisiana Geographic Information Center — This is a point dataset representing 10,044 sequential mile positions of navigable inland waterways and the Gulf Intracoastal Waterway. These data were developed as...

The influence of hydrology and waterway distance on population structure of Chinook salmon Oncorhynchus tshawytscha in a large river.

Science.gov (United States)

Olsen, J B; Beacham, T D; Wetklo, M; Seeb, L W; Smith, C T; Flannery, B G; Wenburg, J K

2010-04-01

Adult Chinook salmon Oncorhynchus tshawytscha navigate in river systems using olfactory cues that may be influenced by hydrologic factors such as flow and the number, size and spatial distribution of tributaries. Thus, river hydrology may influence both homing success and the level of straying (gene flow), which in turn influences population structure. In this study, two methods of multivariate analysis were used to examine the extent to which four indicators of hydrology and waterway distance explained population structure of O. tshawytscha in the Yukon River. A partial Mantel test showed that the indicators of hydrology were positively associated with broad-scale (Yukon basin) population structure, when controlling for the influence of waterway distance. Multivariate multiple regression showed that waterway distance, supplemented with the number and flow of major drainage basins, explained more variation in broad-scale population structure than any single indicator. At an intermediate spatial scale, indicators of hydrology did not appear to influence population structure after accounting for waterway distance. These results suggest that habitat changes in the Yukon River, which alter hydrology, may influence the basin-wide pattern of population structure in O. tshawytscha. Further research is warranted on the role of hydrology in concert with waterway distance in influencing population structure in Pacific salmon.

Order of 21 May 1979 on Transport of Radioactive Materials by Road or by Inland Waterway (Dangerous Goods 1979, No.5)

International Nuclear Information System (INIS)

1979-01-01

This Order lays down in connection with radioactive materials (class IVb) to be carried by road or inland waterway that Section 32 of the Order of 15 April 1945 approving the transport of dangerous goods by rail, land or inland waterway will come into force on 1 October 1980. The Order prescribes that drivers of land vehicles and on inland waterways, persons responsible on board must carry a certificate concerning their cargo in accordance with the above Section. (NEA) [fr

Aspects of the state safety regulation dealing with management of radioactive wastes from nuclear vessels

International Nuclear Information System (INIS)

Markarov, Valentin G.

1999-01-01

According to this presentation, the Constitution of the Russian Federation states that nuclear power engineering and fissile materials are under the jurisdiction of the Russian Federation. But there is no federal law with detailed directions for radioactive waste (RW) management, which thus comes under the Federal law ''On Use of Atomic Energy''. This law defines the legal basis and principles of regulating the relations occurring during RW management and sets some general requirements. RW management safety is regulated by the federal norms and rules (1) Radiation Safety Norms (NRB-96), Basic Sanitary Rules (OSP-72, 87) and (3) Sanitary Rules for RW Management (SPORO-85), etc. A number of normative documents on RW management will be put in force in 1999. For work in the field of RW management, licence must in general be obtained from Gozatomnazdor of Russia. The conditions for receiving a license for the management of RW from vessels are presented

2013 West Coast Vessel Tracklines

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

2011 West Coast Vessel Tracklines

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

2013 Great Lakes Vessel Tracklines

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

2011 East Coast Vessel Tracklines

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

2011 Great Lakes Vessel Tracklines

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

ADA access to passenger vessels : finding safety equivalence solutions for weathertight doors with coamings : Phase 2 : a risk management approach to reconfiguration design solutions

Science.gov (United States)

2005-03-01

This report examines a risk management methodology to provide for both marine safety and disability access at weathertight doors into passenger accommodation spaces on U.S. passenger vessels. The Architectural and Transportation Barriers Compliance B...

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

Analyses for passive safety of fusion reactor during ex-vessel loss of coolant accident

International Nuclear Information System (INIS)

Honda, Takuro; Okazaki, Takashi; Maki, Koichi; Uda, Tatuhiko; Seki, Yasushi; Aoki, Isao; Kunugi, Tomoaki.

1995-01-01

Passive safety of nuclear fusion reactors during ex-vessel Loss-of-Coolant Accidents (LOCAs) in the divertor cooling system has been investigated using a hybrid code, which can treat the interaction of the plasma and plasma facing components (PFCs). The code has been modified to include the impurity emission from PFCs with a diffusion model at the edge plasma. We assumed an ex-vessel LOCA of the divertor cooling system during the ignited operation in International Thermonuclear Experimental Reactor (ITER), in which a carbon-copper brazed divertor plate was employed in the Conceptual Design Activity (CDA). When a double-ended break occurs at the cold leg of the divertor cooling system, the impurity density in the main plasma becomes about twice within 2s after the LOCA due to radiation enhanced sublimation of graphite PFCs. The copper cooling tube of the divertor begins to melt at about 3s after the LOCA, even though the plasma is passively shut down at about 4s due to the impurity accumulation. It is necessary to apply other PFC materials, which can shorten the time period for passive shutdown, or an active shutdown system to keep the reactor structures intact for such rapid transient accident. (author)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

33 CFR 165.T09-1080 - Safety Zone and Regulated Navigation Area, Chicago Sanitary and Ship Canal, Romeoville, IL.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety Zone and Regulated Navigation Area, Chicago Sanitary and Ship Canal, Romeoville, IL. 165.T09-1080 Section 165.T09-1080 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PORTS AND WATERWAYS SAFETY REGULATED NAVIGATION AREAS AND LIMITED...

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

First insights into the functional role of vasicentric tracheids and parenchyma in eucalyptus species with solitary vessels: do they contribute to xylem efficiency or safety?

Science.gov (United States)

Barotto, Antonio José; Fernandez, María Elena; Gyenge, Javier; Meyra, Ariel; Martinez-Meier, Alejandro; Monteoliva, Silvia

2016-12-01

The relationship between hydraulic specific conductivity (k s ) and vulnerability to cavitation (VC) with size and number of vessels has been studied in many angiosperms. However, few of the studies link other cell types (vasicentric tracheids (VT), fibre-tracheids, parenchyma) with these hydraulic functions. Eucalyptus is one of the most important genera in forestry worldwide. It exhibits a complex wood anatomy, with solitary vessels surrounded by VT and parenchyma, which could serve as a good model to investigate the functional role of the different cell types in xylem functioning. Wood anatomy (several traits of vessels, VT, fibres and parenchyma) in conjunction with maximum k s and VC was studied in adult trees of commercial species with medium-to-high wood density (Eucalyptus globulus Labill., Eucalyptus viminalis Labill. and Eucalyptus camaldulensis Dehnh.). Traits of cells accompanying vessels presented correlations with functional variables suggesting that they contribute to both increasing connectivity between adjacent vessels-and, therefore, to xylem conduction efficiency-and decreasing the probability of embolism propagation into the tissue, i.e., xylem safety. All three species presented moderate-to-high resistance to cavitation (mean P 50 values = -2.4 to -4.2 MPa) with no general trade-off between efficiency and safety at the interspecific level. The results in these species do not support some well-established hypotheses of the functional meaning of wood anatomy. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please e-mail: journals.permissions@oup.com.

Ships in an Artificial Force Field : A Multi-agent System for Nautical Traffic and Safety

NARCIS (Netherlands)

Xiao, F.

2014-01-01

The main objective of this research is developing a simulation tool that provides information of detailed ship behavior in a specific navigational environment, on both the ship traffic level and the individual ship level, for safety analysis, decision making, planning of ports and waterways, and

DROUGHTS IN THE TIETÊ-PARANÁ WATERWAY: IMPACTS ON THE DIRECT, INDIRECT AND HIDDEN COSTS IN THE TRANSPORTATION OF SOYBEAN.

Directory of Open Access Journals (Sweden)

Rodrigo Carlo Toloi

2016-05-01

Full Text Available Brazil's agricultural economy is growing and increasing productivity. Therefore, it has required transportation systems with high load capacity and lower transportation costs. However, with the drought in the Southeast region of Brazil, the waterway Tietê-Paraná closed since May 2014 generating a loss of more than 30 million last year. Thus, this study investigates the impacts on direct, indirect and hidden costs resulting from this change of route for soy transport. The methodology consists of an exploratory, descriptive and bibliographic research that seeks to raise the main costs. The results show that failing to ensure the production of soybeans by the Tiete-Parana waterway and using the highway transportation costs for waterway users are increased by US$ 37,760,146.86.

33 CFR 165.1401 - Apra Harbor, Guam-safety zones.

Science.gov (United States)

2010-07-01

... (CONTINUED) PORTS AND WATERWAYS SAFETY REGULATED NAVIGATION AREAS AND LIMITED ACCESS AREAS Specific Regulated...′47″ N and 144°39′01.9″ E. Based on World Geodetic System 1984 Datum) (b) The following is designated... the center of Naval Wharf Kilo. (Located at 13°26′43″ N, 144°37′46.7″ E. Based on World Geodetic...

Public safety around dams

Energy Technology Data Exchange (ETDEWEB)

Bourassa, H. [Centre d' expertise hydrique du Quebec, Quebec, PQ (Canada)

2009-07-01

Fourty public dams are managed on a real-time basis by the Centre d'expertise hydrique du Quebec (CEHQ). This presentation described the public dams owned by the CEHQ and discussed the public safety measures at the dams. The dams serve various purposes, including protection against floods; industrial or drinking water supply; resort or recreational activities; hydroelectric development; and wildlife conservation. Trigger events were also discussed, such as the complaint at Rapides-des-Cedres dam and deaths that occurred in 2004 when water from a dam was released without warning. Several photographs were presented to illustrate that people were unaware of the danger. Initiatives aimed at raising awareness and studying public safety issues were discussed. A pilot project was launched and a permanent committee was created to evaluate all aspects of public safety at the dams owned by CEHQ. The first tasks of the committee were to establish requirements for waterway safety barriers, both upstream and downstream, for all public dams; to establish requirements for safety signage for all public dams; and to develop criteria to decide on safety signage at each dam. figs.

Public safety around dams

Energy Technology Data Exchange (ETDEWEB)

Bourassa, H [Centre d' expertise hydrique du Quebec, Quebec, PQ (Canada)

2009-07-01

Fourty public dams are managed on a real-time basis by the Centre d'expertise hydrique du Quebec (CEHQ). This presentation described the public dams owned by the CEHQ and discussed the public safety measures at the dams. The dams serve various purposes, including protection against floods; industrial or drinking water supply; resort or recreational activities; hydroelectric development; and wildlife conservation. Trigger events were also discussed, such as the complaint at Rapides-des-Cedres dam and deaths that occurred in 2004 when water from a dam was released without warning. Several photographs were presented to illustrate that people were unaware of the danger. Initiatives aimed at raising awareness and studying public safety issues were discussed. A pilot project was launched and a permanent committee was created to evaluate all aspects of public safety at the dams owned by CEHQ. The first tasks of the committee were to establish requirements for waterway safety barriers, both upstream and downstream, for all public dams; to establish requirements for safety signage for all public dams; and to develop criteria to decide on safety signage at each dam. figs.

2013 Gulf of Mexico Vessel Tracklines

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

2011 Gulf of Mexico Vessel Tracklines

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

2011 Pleasure Craft Sailing Vessel Density

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

New method of safety assessment for pressure vessel of nuclear power plant--brief introduction of master curve approach

International Nuclear Information System (INIS)

Yang Wendou

2011-01-01

The new Master Curve Method is called as a revolutionary advance to the assessment of- reactor pressure vessel integrity in USA. This paper explains the origin, basis and standard of the Master Curve from the reactor pressure-temperature limit curve which assures the safety of nuclear power plant. According to the characteristics of brittle fracture which is greatly susceptible to the microstructure, the theory and the test method of the Master Curve as well as its statistical law which can be modeled using Weibull distribution are described in this paper. The meaning, advantage, application and importance of the Master Curve as well as the relation between the Master Curve and nuclear power safety are understood from the fitting formula for the fracture toughness database by Weibull distribution model. (author)

75 FR 61094 - Drawbridge Operation Regulation; Atlantic Intracoastal Waterway, Beaufort, SC

Science.gov (United States)

2010-10-04

.... Highway 21, Sea Island Parkway Bridge at mile marker 536 over the Beaufort River (Atlantic Intracoastal... Bridge, which is across the Atlantic Intracoastal Waterway at mile 539.0 in Port Royal, South Carolina... Bridge, across the Beaufort River, Mile 536.0 at Beaufort. The draw shall operate as follows: (1) On...

A history of trade routes and water-level regulation on waterways in Voyageurs National Park, Minnesota, USA

Science.gov (United States)

Christensen, Victoria G.; LaBounty, Andrew E.

2018-01-01

Unlike most national parks, main access to Voyageurs National Park is by boat. This remote system of interconnected waterways along the USA-Canada border was an important transportation route for thousands of years of American Indian occupation, leading up to and including the trade route of the voyageurs, or French-Canadian fur traders from around 1680 to 1870. The Ojibwe people collaborated with the voyageurs and the two cultures developed a trade network that continued to rely on these waterways. By the mid-1800s, European fashion changed, and the fur trade dwindled while the Ojibwe remained tied to the land and waters. The complexity of the waterways increased with the installation of dams on two of the natural lakes in the early 1900s. Modern water levels have affected—and in some cases destabilized—vulnerable landforms within the past century. The knowledge of these effects can be used by resource managers to weigh the consequences of hydrologic manipulation in Voyageurs National Park.

Safety valve opening and closing operation monitor

International Nuclear Information System (INIS)

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

1981-01-01

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

On the Waterway Staff Training Management and Innovation%试论航道员工培训管理与创新

Institute of Scientific and Technical Information of China (English)

镇卫兵

2014-01-01

近年来我国航道事业随着社会经济的快速发展越来越被重视，航道作为水路运输的重要组成部分，其管理状况对水运发展起着直接影响作用。本文通过对航道现行体制与管理情况探讨，结合员工培训管理与创新分析，提出加强航道管理水平的策略，以实现水运交通事业快速持续健康发展的目标。%In recent years, with the rapid development of social economy, waterway business in China has been paid more and more attention. As an important part of water transportation, waterway's management conditions have a direct impact on water transportation development. Through the discussion of current system and management conditions of waterway, combined with analysis of employee training management and innovation, this article puts forward the strategies to strengthen the waterway management level, in order to achieve the goal of rapid, sustainable and healthy development of waterway transportation business.

Thermal radiation from fireballs on failure of liquefied petroleum gas storage vessels

Energy Technology Data Exchange (ETDEWEB)

Roberts, T.; Hawksworth, S. [Health and Safety Executive, Health and Safety Lab., Buxton (United Kingdom); Gosse, A. [BG Technology, Loughborough (United Kingdom)

2000-05-01

Fire impingement on vessels containing pressure liquefied gases can result in catastrophic failure of the vessel leading to a Boiling Liquid Expanding Vapour Explosion (BLEVE). If the gas is flammable, this can result in the formation of very large fireballs. In safety assessments where catastrophic vessel failure is identified as a real possibility, the risk of death from a fireball tends to be higher than that from missiles or blast. Since many of the physical processes which take place in a BLEVE are scale dependent, a series of tests were undertaken at a large scale where 2 tonne propane vessels were taken to failure in a jet fire and the vessel response, mode of failure and consequence of failure characterised. The measurements taken by the Health and Safety Laboratory and BG Technology relating to fireball formation are described. (Author)

Efficacy of commercial algaecides to manage species of Phytophthora in suburban waterways

Science.gov (United States)

G. Curtis Colburn; Steven N. Jeffers

2010-01-01

Many commercial algaecides contain copper compounds as active ingredients. Phytophthora spp. and other oomycetes are known to be sensitive to copper-based fungicides. Therefore, algaecides registered to manage algae in natural waterways and irrigation waters also might be effective for mitigating or even eradicating Phytophthora ...

33 CFR 96.220 - What makes up a safety management system?

Science.gov (United States)

2010-07-01

... system? 96.220 Section 96.220 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY VESSEL OPERATING REGULATIONS RULES FOR THE SAFE OPERATION OF VESSELS AND SAFETY MANAGEMENT SYSTEMS Company and Vessel Safety Management Systems § 96.220 What makes up a safety management system? (a) The...

Effectiveness of In-Vessel Retention Strategies and Minimum Safety Injection Flow over Postulated Severe Accidents of OPR1000

International Nuclear Information System (INIS)

Kim, Sung Joong; Seo, Seungwon; Lee, Seongnyeon; KIm, Hwan Yeol; Ha, Kwang Soon; Park, Jonghwa; Park, Raejoon

2013-01-01

The objective of this study is first to evaluate various serious severe accident scenarios of OPR1000 with and without in-vessel retention strategies using MELCOR code. Second is to develop a mechanistic model of minimum safety injection flow using the thermal-hydraulic parameters of CET and collapsed water level obtained from the MELCOR simulation results. Effectiveness of RCS depressurization of OPR1000 is investigated for postulated severe accidents of SBLOCA, SBO, and TLOF. It is seen that timely operator action is important to achieve the best mitigation. Also The MELCOR simulation results of SBLOCA, SBO, and TLOFW are utilized to develop a model for minimum safety injection flow. The model suggests that if HPSI is available with RCS pressure lower than 120 bars, the core coolability can be guaranteed. In this study, several MELCOR simulations are conducted in search for effective in-vessel retention strategies over postulated severe accidents of SBLOCA, SBO, and TLOFW of OPR1000. Detailed accident sequences are presented and indicative parameters diagnosing the reactor thermal-hydraulic state are interrogated to provide useful information to the operator actions. To properly assist operator's action during the severe accident, the thermal-hydraulic parameters should be virtual, intuitive, and reliable. In addition, the parameters should be collected through the instrumentations close to the reactor core. In this regard, Core Exit Temperature (CET) and collapsed core water level are deemed as the commensurate parameters

Effectiveness of In-Vessel Retention Strategies and Minimum Safety Injection Flow over Postulated Severe Accidents of OPR1000

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung Joong; Seo, Seungwon; Lee, Seongnyeon [Hanyang Univ., Seoul (Korea, Republic of); KIm, Hwan Yeol; Ha, Kwang Soon; Park, Jonghwa; Park, Raejoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

The objective of this study is first to evaluate various serious severe accident scenarios of OPR1000 with and without in-vessel retention strategies using MELCOR code. Second is to develop a mechanistic model of minimum safety injection flow using the thermal-hydraulic parameters of CET and collapsed water level obtained from the MELCOR simulation results. Effectiveness of RCS depressurization of OPR1000 is investigated for postulated severe accidents of SBLOCA, SBO, and TLOF. It is seen that timely operator action is important to achieve the best mitigation. Also The MELCOR simulation results of SBLOCA, SBO, and TLOFW are utilized to develop a model for minimum safety injection flow. The model suggests that if HPSI is available with RCS pressure lower than 120 bars, the core coolability can be guaranteed. In this study, several MELCOR simulations are conducted in search for effective in-vessel retention strategies over postulated severe accidents of SBLOCA, SBO, and TLOFW of OPR1000. Detailed accident sequences are presented and indicative parameters diagnosing the reactor thermal-hydraulic state are interrogated to provide useful information to the operator actions. To properly assist operator's action during the severe accident, the thermal-hydraulic parameters should be virtual, intuitive, and reliable. In addition, the parameters should be collected through the instrumentations close to the reactor core. In this regard, Core Exit Temperature (CET) and collapsed core water level are deemed as the commensurate parameters.

Long-term safety and feasibility of three-vessel multimodality intravascular imaging in patients with ST-elevation myocardial infarction

DEFF Research Database (Denmark)

Taniwaki, Masanori; Radu, Maria D; Garcia-Garcia, Hector M

2015-01-01

We assessed the feasibility and the procedural and long-term safety of intracoronary (i.c) imaging for documentary purposes with optical coherence tomography (OCT) and intravascular ultrasound (IVUS) in patients with acute ST-elevation myocardial infarction (STEMI) undergoing primary PCI in the s......We assessed the feasibility and the procedural and long-term safety of intracoronary (i.c) imaging for documentary purposes with optical coherence tomography (OCT) and intravascular ultrasound (IVUS) in patients with acute ST-elevation myocardial infarction (STEMI) undergoing primary PCI...... in the setting of IBIS-4 study. IBIS4 (NCT00962416) is a prospective cohort study conducted at five European centers including 103 STEMI patients who underwent serial three-vessel coronary imaging during primary PCI and at 13 months. The feasibility parameter was successful imaging, defined as the number...... of pullbacks suitable for analysis. Safety parameters included the frequency of peri-procedural complications, and major adverse cardiac events (MACE), a composite of cardiac death, myocardial infarction (MI) and any clinically-indicated revascularization at 2 years. Clinical outcomes were compared...

78 FR 66266 - Drawbridge Operation Regulation; Atlantic Intracoastal Waterway (AIWW), Chesapeake, VA

Science.gov (United States)

2013-11-05

... deviation from the operating schedule that governs the I64 Bridge across the Atlantic Intracoastal Waterway... deviation, call or email Mrs. Kashanda Booker, Bridge Administration Branch Fifth District, Coast Guard... maintenance of the moveable spans on the structure. The current operating schedule for the drawbridge is set...

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)

Sailing Vessel Routing Considering Safety Zone and Penalty Time for Altering Course

Directory of Open Access Journals (Sweden)

Marcin Zyczkowski

2017-06-01

Full Text Available In this paper we introduce new model for simulation sea vessel routing. Besides a vessel types (polar diagram and weather forecast, travel security and the number of maneuvers are considered. Based on these data both the minimal travelling costs and the minimal processing time are found for different vessels and different routes. To test our model the applications SailingAssistance wad improved. The obtained results shows that we can obtain quite acceptable results.

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)

46 CFR 50.05-5 - Existing boilers, pressure vessels or piping systems.

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Existing boilers, pressure vessels or piping systems. 50... ENGINEERING GENERAL PROVISIONS Application § 50.05-5 Existing boilers, pressure vessels or piping systems. (a) Whenever doubt exists as to the safety of an existing boiler, pressure vessel, or piping system, the marine...

33 CFR 96.370 - What are the requirements for vessels of countries not party to Chapter IX of SOLAS?

Science.gov (United States)

2010-07-01

... vessel, or self-propelled mobile offshore drilling unit of 500 gross tons or more, operated in U.S... Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY VESSEL OPERATING REGULATIONS RULES FOR THE SAFE OPERATION OF VESSELS AND SAFETY MANAGEMENT SYSTEMS How Will Safety Management Systems Be Certificated and...

33 CFR 96.230 - What objectives must a safety management system meet?

Science.gov (United States)

2010-07-01

... management system meet? 96.230 Section 96.230 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY VESSEL OPERATING REGULATIONS RULES FOR THE SAFE OPERATION OF VESSELS AND SAFETY MANAGEMENT SYSTEMS Company and Vessel Safety Management Systems § 96.230 What objectives must a safety...

2013 Pleasure Craft and Sailing Vessel Density

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

Recreation value of a waterway and changes in this value

International Nuclear Information System (INIS)

Aittoniemi, P.

1991-01-01

The uniquely rich hydrological and archipelagic nature of Finland creates versatile opportunities for spending of leisure time. As various uses of waterways compete with each other, it is difficult to compare advantages and disadvantages of operations in view of various interests. This is why efforts have been made to develop methods to make the effect commensurable. Usability of shores is one important factor in the assessment of recreation use of waterways. When influence of water level changes in usability of shores is estimated, shore and bottom slope and quality, and uses and user numbers of the shores must be known. In several assessment methods, usability of shore is depicted by distance of the shifted shoreline (due to the changed water level) from the 'good' status. The studies on changes in the recreation value or usability of shores should be started by making efforts to classify and estimate various effects separately. If the value is wished to be given as one figure, the calculation basis applied, the weights and the used value functions must be stated clearly with the results. It is equally important to make a sensitivity analysis, i.e. to study effects of changes in uncertain input data on the results

Research and development of the prestressed concrete reactor vessel

International Nuclear Information System (INIS)

Shiozawa, Shoji; Omata, Ippei; Nakamura, Norio

1975-01-01

Compared with the steel reactor vessel, the prestressed concrete reactor vessel (PCRV) is said to be superior in safety and economy. One of the characteristics of the high temperature gas cooled reactor (HTGR) is the adoption of the PCRV instead of the steel reactor vessel to ensure safety. In order to improve safety characteristics, it is necessary for the PCRV to be provided with more reliable functions. When the multi-purpose HTGR or the gas cooled fast breeder reactor (GCFR) are realized in future, more severe conditions of technology will be imposed on the PCRV, and accordingly, technical developments are now increasingly required. IHI is now proceeding with the technical research and development on the PCRV, in which a basic study of its liner cooling system has already been completed. In this study applying a large cylindrical PCRV model, comparison was made between experimental data and analyses concerning the liner cooling system, and the results of analytical technique have been evaluated. The analytical technique established this time is applicable to the estimation of temperature distribution in the concrete of a large PCRV and also to the evaluation of the liner cooling system. (auth.)

33 CFR 96.240 - What functional requirements must a safety management system meet?

Science.gov (United States)

2010-07-01

... a safety management system meet? 96.240 Section 96.240 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY VESSEL OPERATING REGULATIONS RULES FOR THE SAFE OPERATION OF VESSELS AND SAFETY MANAGEMENT SYSTEMS Company and Vessel Safety Management Systems § 96.240 What functional...

Failure probability analysis on mercury target vessel

International Nuclear Information System (INIS)

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

2005-03-01

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

Design of pressure vessels. Part 1

International Nuclear Information System (INIS)

Grandemange, J.M.

2008-01-01

The equipments and loops of PWR reactors are basically pressure vessels. Their specificities concern the integrity warranties that must be implemented considering their importance for the reactors safety. Thus, stress is put on the exhaustiveness of the prevention of in-service degradation and on the safety scenarios considered. The second specificity concerns the possibility of activation of wear and corrosion products during their flow inside the reactor core. This second aspect leads to some constraints on the choice of the materials used and on the surface coating of the inside wall of big components of the primary circuit. The aim of this document is to develop the general approach adopted for the design of the pressure vessels of PWR fluid loops, and to stress more particularly on the nuclear particularities of these equipments. Some extensions of these rules to high temperature resistant materials (FBR-type reactors) are also evoked. Content: General considerations: design basis of pressure vessels, risk analysis and design conditions, ruining paths and safety coefficients; 2 - damage prevention for excessive deformation: definitions, criteria; 3 - prevention of the plastic instability damage: definition, criteria; 4 - buckling prevention: definition and mechanisms, rules and criteria; 5 - prevention of progressive deformation damage: definitions, plastic adaptation, plastic accommodation, progressive deformation; 6 - prevention of fatigue damage: definitions, general prevention approach, design fatigue curves, analytic approach, particular aspects, analysis of zones with geometrical singularity; 7 - prevention of sudden rupture damage: fragile rupture and ductile tear, general approach, analytic criteria, irradiation and aging effects; 8 - other potential damages; 9 - conclusion. (J.S.)

Safety Evaluation Report: Development of Improved Composite Pressure Vessels for Hydrogen Storage, Lincoln Composites, Lincoln, NE, May 25, 2010

Energy Technology Data Exchange (ETDEWEB)

Fort, III, William C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kallman, Richard A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maes, Miguel [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Skolnik, Edward G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Weiner, Steven C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2010-12-22

Lincoln Composites operates a facility for designing, testing, and manufacturing composite pressure vessels. Lincoln Composites also has a U.S. Department of Energy (DOE)-funded project to develop composite tanks for high-pressure hydrogen storage. The initial stage of this project involves testing the permeation of high-pressure hydrogen through polymer liners. The company recently moved and is constructing a dedicated research/testing laboratory at their new location. In the meantime, permeation tests are being performed in a corner of a large manufacturing facility. The safety review team visited the Lincoln Composites site on May 25, 2010. The project team presented an overview of the company and project and took the safety review team on a tour of the facility. The safety review team saw the entire process of winding a carbon fiber/resin tank on a liner, installing the boss and valves, and curing and painting the tank. The review team also saw the new laboratory that is being built for the DOE project and the temporary arrangement for the hydrogen permeation tests.

Controls on denitrification potential in nitrate-rich waterways and riparian zones of an irrigated agricultural setting.

Science.gov (United States)

Webster, Alex J; Groffman, Peter M; Cadenasso, Mary L

2018-02-21

Denitrification, the microbial conversion of NO 3 - to N gases, is an important process contributing to whether lotic and riparian ecosystems act as sinks for excess NO 3 - from agricultural activities. Though agricultural waterways and riparian zones have been a focus of denitrification research for decades, almost none of this research has occurred in the irrigated agricultural settings of arid and semi-arid climates. In this study, we conducted a broad survey of denitrification potential in riparian soils and channel sediment from 79 waterway reaches in the irrigated agricultural landscape of California's Central Valley. With this approach, we sought to capture the wide range of variation that arose from diverse waterway management and fluctuating flow conditions, and use this variation to identify promising management interventions. We explored associations of denitrification potentials with surface water NO 3 - -N, organic matter, flow conditions, vegetation cover, near-channel riparian bank slope, and channel geomorphic features using generalized linear mixed models. We found strong associations of sediment denitrification potentials with reach flow conditions, which we hypothesize was the result of variation in microbial communities' tolerance to dry-wet cycles. Denitrification potentials in riparian soils, in contrast, did not appear affected by flow conditions, but instead were associated with organic matter, vegetation cover, and bank slope in the riparian zone. These results suggest a strong need for further work on how denitrification responds to varying flow conditions and dry-wet cycles in non-perennial lotic ecosystems. Our findings also demonstrate that denitrifier communities respond to key features of waterway management, which can therefore be leveraged to control denitrification through a variety of management actions. This article is protected by copyright. All rights reserved. This article is protected by copyright. All rights reserved.

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.

Effectiveness of maritime safety control in different navigation zones using a spatial sequential DEA model: Yangtze River case.

Science.gov (United States)

Wu, Bing; Wang, Yang; Zhang, Jinfen; Savan, Emanuel Emil; Yan, Xinping

2015-08-01

This paper aims to analyze the effectiveness of maritime safety control from the perspective of safety level along the Yangtze River with special considerations for navigational environments. The influencing variables of maritime safety are reviewed, including ship condition, maritime regulatory system, human reliability and navigational environment. Because the former three variables are generally assumed to be of the same level of safety, this paper focuses on studying the impact of navigational environments on the level of safety in different waterways. An improved data envelopment analysis (DEA) model is proposed by treating the navigational environment factors as inputs and ship accident data as outputs. Moreover, because the traditional DEA model cannot provide an overall ranking of different decision making units (DMUs), the spatial sequential frontiers and grey relational analysis are incorporated into the DEA model to facilitate a refined assessment. Based on the empirical study results, the proposed model is able to solve the problem of information missing in the prior models and evaluate the level of safety with a better accuracy. The results of the proposed DEA model are further compared with an evidential reasoning (ER) method, which has been widely used for level of safety evaluations. A sensitivity analysis is also conducted to better understand the relationship between the variation of navigational environments and level of safety. The sensitivity analysis shows that the level of safety varies in terms of traffic flow. It indicates that appropriate traffic control measures should be adopted for different waterways to improve their safety. This paper presents a practical method of conducting maritime level of safety assessments under dynamic navigational environment. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

Fatal occupational accidents in Danish fishing vessels 1989-2005

DEFF Research Database (Denmark)

Laursen, Lise Hedegaard; Hansen, Henrik L; Jensen, Olaf

2008-01-01

training for all fishermen and improved safety measures are needed, especially in the underscored areas of sea disasters concerning small vessels and occupational accidents on big vessels. Better registration of time at risk for fishermen is needed to validate the effect of the safety measures......./capsizing due to stability changes in rough weather and collisions; 39 fatal occupational accidents mainly occurred on the larger inspection obligated trawlers during fishing. In the remaining 14 other fatal accidents, the main causal factors were difficult embarking/disembarking conditions by darkness...... in foreign ports and alcohol intoxication. In the period 1995-2005, the overall incidence rate was 10 per 10,000 fishermen per year with no down-going trend during that period. The fatal accident rates are still too high, despite the efforts to reduce the risk. Increased focus on regular and repeated safety...

Management of Environmental Resources of Cutoff Bends Along the Tennessee-Tombigbee Waterway.

Science.gov (United States)

1987-11-01

WW2 , pp 51-67. Carey, W. C. 1964. "Discussion of ’Hydraulic Aspects of Arkansas River Sta- bilization,’" Journal of the Waterways and Harbors Division...photographs listed in Table A6, color infrared (IR) photographs of several of the Tombigbee River sites were obtained and used in a qualitative fashion. For

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)

Emergency venting of pressure vessels

International Nuclear Information System (INIS)

Steinkamp, H.

1995-01-01

With the numerical codes developed for safety analysis the venting of steam vessel can be simulated. ATHLET especially is able to predict the void fraction depending on the vessel height. Although these codes contain a one-dimensional model they allow the description of complex geometries due to the detailed nodalization of the considered apparatus. In chemical reactors, however, the venting process is not only influenced by the flashing behaviour but additionally by the running chemical reaction in the vessel. Therefore the codes used for modelling have to consider the kinetics of the chemical reaction. Further multi-component systems and dissolving processes have to be regarded. In order to preduct the fluid- and thermodynamic process it could be helpful to use 3-dimensional codes in combination with the one-dimensional codes as used in nuclear industry to get a more detailed describtion of the running processes. (orig./HP)

75 FR 33690 - Safety Zone, Lights on the River Fireworks Display, Delaware River, New Hope, PA

Science.gov (United States)

2010-06-15

... scenario with potential for loss of life and property. Basis and Purpose The New Hope Chamber of Commerce... to protect life and property operating on the navigable waterways of the Delaware River in New Hope...-AA00 Safety Zone, Lights on the River Fireworks Display, Delaware River, New Hope, PA AGENCY: Coast...

Generic analyses for evaluation of low Charpy upper-shelf energy effects on safety margins against fracture of reactor pressure vessel materials

International Nuclear Information System (INIS)

Dickson, T.L.

1993-07-01

Appendix G to 10 CFR Part 50 requires that reactor pressure vessel beltline material maintain Charpy upper-shelf energies of no less than 50 ft-lb during the plant operating life, unless it is demonstrated in a manner approved by the Nuclear Regulatory Commission (NRC), that lower values of Charpy upper-shelf energy provide margins of safety against fracture equivalent to those in Appendix G to Section XI of the ASME Code. Analyses based on acceptance criteria and analysis methods adopted in the ASME Code Case N-512 are described herein. Additional information on material properties was provided by the NRC, Office of Nuclear Regulatory Research, Materials Engineering Branch. These cases, specified by the NRC, represent generic applications to boiling water reactor and pressurized water reactor vessels. This report is designated as HSST Report No. 140

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)

The target programme for water pollution control up to year 2005. The impact to alternative pollution load levels on inland waterways; Vesiensuojelun tavoitteet vuoteen 2005. Vaihtoehtoisten kuormitustasojen vaikutukset sisaevesissae

Energy Technology Data Exchange (ETDEWEB)

Marttunen, M. [eds.

1998-11-01

The publication presents waterway-specific pollution loading surveys connected with the target programme for water pollution control up to 2005 and the results of a survey concerning inland waters. The work evaluates and compares the magnitude and significance of future pollution loads from various sources on the state of the waterway and assesses the impact on the condition of the waterway of two alternative pollution loading levels. The survey was conducted on 24 lakes or parts thereof and 10 rivers. Sixteen of these are subject to loading from the pulls and paper industry. The waterways selected for examination were especially ones for which relatively recent loading analyses were available and to which water quality models have been applied. The general classification of fitness for use was used as a benchmark for changes occurring in the water quality of the waterways. The findings show that particularly the large waterway systems can be moved into a higher fitness for use class by reducing substantially the nutrient loading. Significant improvement in the condition of eutrophic lakes and coastal rivers with their internal loads is difficult. (orig.) 26 refs.

Safety analyses for transient behavior of plasma and in-vessel components during plasma abnormal events in fusion reactor

International Nuclear Information System (INIS)

Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

1997-01-01

Safety analyses on plasma abnormal events have been performed using a hybrid code of a plasma dynamics model and a heat transfer model of in-vessel components. Several abnormal events, e.g., increase in fueling rate, were selected for the International Thermonuclear Experimental Reactor (ITER) and transient behavior of the plasma and the invessel components during the events was analyzed. The physics model for safety analysis was conservatively prepared. In most cases, the plasma is terminated by a disruption or it returns to the original operation point. When the energy confinement improves by a factor of 2.0 in the steady state, which is a hypothetical assumption under the present plasma data, the maximum fusion power reaches about 3.3 GW at about 3.6 s and the plasma is terminated due to a disruption. However, the results obtained in this study show the confinement boundary of ITER can be kept almost intact during the abnormal plasma transients, as long as the cooling system works normally. Several parametric studies are needed to comprehend the overpower transient including structure behavior, since many uncertainties are connected to the filed of the plasma physics. And, future work will need to discuss the burn control scenario considering confinement mode transition, system specifications, experimental plans and safety regulations, etc. to confirm the safety related to the plasma anomaly. (author)

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

International Nuclear Information System (INIS)

Steinwarz, Wolfgang; Bounin, Dieter

2014-01-01

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

33 CFR 96.250 - What documents and reports must a safety management system have?

Science.gov (United States)

2010-07-01

... safety management system have? 96.250 Section 96.250 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY VESSEL OPERATING REGULATIONS RULES FOR THE SAFE OPERATION OF VESSELS AND SAFETY MANAGEMENT SYSTEMS Company and Vessel Safety Management Systems § 96.250 What documents and...

75 FR 29695 - Security Zones; Sabine Bank Channel, Sabine Pass Channel and Sabine-Neches Waterway, TX

Science.gov (United States)

2010-05-27

... rule under Executive Order 13045, Protection of Children from Environmental Health Risks and Safety Risks. This rule is not an economically significant rule and would not create an environmental risk to... proposes to establish moving security zones for certain vessels for which the Captain of the Port, Port...

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....

Resolution of the Task A-11 reactor-vessel materials-toughness safety issue. Part I. Main report. Part II. Staff responses to public comments, and Appendices A and B

International Nuclear Information System (INIS)

Johnson, R.

1982-10-01

This report provides the NRC position with respect to the reactor pressure vessel safety analysis required according to the rules given in the Code of Federal Regulations, Title 10 (10 CFR). An analysis is required whenever neutron irradiation reduces the Charpy V-notch upper shelf energy level in the vessel steel to 50 ft-lb or less. Task A-11 was needed because the available engineering methodology for such an analysis utilized linear elastic fracture mechanics principles, which could not fully account for the plastic deformation or stable crack extension expected at upper shelf temperatures. The Task A-11 goal was to develop an elastic-plastic fracture mechanics methodology, applicable to the beltline region of a pressurized water reactor vessel, which could be used in the required safety analysis. The goal was achieved with the help of a team of recognized experts. Part I of this volume contains the For Comment NUREG-0744, originally published in September 1981 and edited to accommodate comments from the public and the NRC staff. Edited segments are noted by vertical marginal lines. Part II of this volume contains the staff's responses to, and resolution of, the public comments received

Pressure vessel for a BWR type reactor

International Nuclear Information System (INIS)

Shimamoto, Yoshiharu.

1980-01-01

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

Some aspects of reactor pressure vessel integrity

International Nuclear Information System (INIS)

Korosec, D.; Vojvodic, G.J.

1996-01-01

Reactor pressure vessel of the pressurized water reactor nuclear power plant is the subject of extreme interest due to the fact that presents the pressure boundary of the reactor coolant system, which is under extreme thermal, mechanical and irradiation effects. Reactor pressure vessel by itself prevents the release of fission products to the environment. Design, construction and in-service inspection of such component is governed by strict ASME rules and other forms of administrative control. The reactor pressure vessel in nuclear power plant Kriko is designed and constructed in accordance with related ASME rules. The in-service inspection program includes all requests presented in ASME Code section XI. In the present article all major requests for the periodic inspections of reactor pressure vessel and fracture mechanics analysis are discussed. Detailed and strict fulfillment of all prescribed provisions guarantee the appropriate level of nuclear safety. (author)

Evaluation of Agency Non-Code Layered Pressure Vessels (LPVs)

Science.gov (United States)

Prosser, William H.

2014-01-01

In coordination with the Office of Safety and Mission Assurance and the respective Center Pressure System Managers (PSMs), the NASA Engineering and Safety Center (NESC) was requested to formulate a consensus draft proposal for the development of additional testing and analysis methods to establish the technical validity, and any limitation thereof, for the continued safe operation of facility non-code layered pressure vessels. The PSMs from each NASA Center were asked to participate as part of the assessment team by providing, collecting, and reviewing data regarding current operations of these vessels. This report contains the outcome of the assessment and the findings, observations, and NESC recommendations to the Agency and individual NASA Centers.

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

International Nuclear Information System (INIS)

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

1977-01-01

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

76 FR 8653 - Drawbridge Operation Regulation; Gulf Intracoastal Waterway, New Orleans Harbor, Inner Harbor...

Science.gov (United States)

2011-02-15

... Operation Regulation; Gulf Intracoastal Waterway, New Orleans Harbor, Inner Harbor Navigation Canal, New Orleans, Orleans Parish, LA AGENCY: Coast Guard, DHS. ACTION: Notice of temporary deviation from... Lock), at New Orleans, Orleans Parish, Louisiana. This deviation is necessary to replace all of the...

75 FR 78601 - Drawbridge Operation Regulation; Gulf Intracoastal Waterway, New Orleans Harbor, Inner Harbor...

Science.gov (United States)

2010-12-16

... Operation Regulation; Gulf Intracoastal Waterway, New Orleans Harbor, Inner Harbor Navigation Canal, New Orleans, Orleans Parish, LA AGENCY: Coast Guard, DHS. ACTION: Notice of temporary deviation from... Harvey Lock), at New Orleans, Orleans Parish, Louisiana. This deviation is necessary to adjust the...

46 CFR 91.60-5 - Cargo Ship Safety Construction Certificate.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Cargo Ship Safety Construction Certificate. 91.60-5... VESSELS INSPECTION AND CERTIFICATION Certificates Under International Convention for Safety of Life at Sea, 1974 § 91.60-5 Cargo Ship Safety Construction Certificate. (a) All vessels on an international voyage...

46 CFR 189.60-10 - Cargo Ship Safety Equipment Certificate.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Cargo Ship Safety Equipment Certificate. 189.60-10... VESSELS INSPECTION AND CERTIFICATION Certificates Under International Convention for Safety of Life at Sea, 1974 § 189.60-10 Cargo Ship Safety Equipment Certificate. (a) All vessels on an international voyage...

46 CFR 189.60-5 - Cargo Ship Safety Construction Certificate.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Cargo Ship Safety Construction Certificate. 189.60-5... VESSELS INSPECTION AND CERTIFICATION Certificates Under International Convention for Safety of Life at Sea, 1974 § 189.60-5 Cargo Ship Safety Construction Certificate. (a) All vessels on an international voyage...

Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessels

International Nuclear Information System (INIS)

2005-10-01

. The guidance reports are directed at technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant components addressed in the reports. This report addresses the reactor pressure vessel (RPV) in BWRs. Maintaining the structural integrity of this RPV throughout NPP service life in spite of several ageing mechanisms is essential for plant safety

Position paper: Inland waterways classification for South America : Core concepts and initial proposals

NARCIS (Netherlands)

Jaimurzina, A.; Koedijk, O.C.; Wilmsmeier, G.; Dohms, A.; Montiel, D.; Pauli, G.; Rigo, P; Spengler, T.; Wens, F.

2016-01-01

South America, as of yet, has not been able to take full advantage of its extensive system of naturally navigable waterways and in making them an integrated part of the region’s transport network to cater for the ever increasing demand for cargo and human mobility. Infrastructure limitations are one

Assessment and management of ageing of major nuclear power plant components important to safety: PWR vessel internals

International Nuclear Information System (INIS)

1999-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed at technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant

46 CFR 26.03-1 - Safety orientation.

Science.gov (United States)

2010-10-01

... 46 Shipping 1 2010-10-01 2010-10-01 false Safety orientation. 26.03-1 Section 26.03-1 Shipping... Requirements § 26.03-1 Safety orientation. (a) Before getting underway on any uninspected passenger vessel, the... this subpart engaged in tender service at yacht clubs and marinas, and vessels being demonstrated for a...

Design of the ITER vacuum vessel

International Nuclear Information System (INIS)

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

1995-01-01

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

Information About Dynamics of the Sea Surface as a Means to Improve Safety of the Unmanned Vessel at Sea

Directory of Open Access Journals (Sweden)

Przyborski Marek

2016-12-01

Full Text Available One of the fundamental states of the sea surface is its heave. Despite of years of the intense scientific inquiry, no clear understanding of the influence of this aspect on the dynamics of the sea environment has emerged. The separation of two nearby fluid elements which one may observed for example as a free floating of small objects on the sea surface (rescuers on the rough sea or small research vessels is caused by the interaction of different components. On the other hand one may say that the heave of the sea is also a summary interaction of a few components describing the dynamics of the sea. Therefore it is the most important aspect, which influenced the dispersion phenomenon. This observation has important consequences for many different problems as for example conducting Search and Rescue missions and using unmanned ships. We would like to present results of our experiment focused on finding the answer to question about nature of the heave of the sea and its influence on safety of Unmanned Surface Vessels (USV.

46 CFR 189.60-15 - Cargo Ship Safety Radio Certificate.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Cargo Ship Safety Radio Certificate. 189.60-15 Section... VESSELS INSPECTION AND CERTIFICATION Certificates Under International Convention for Safety of Life at Sea, 1974 § 189.60-15 Cargo Ship Safety Radio Certificate. Every vessel equipped with a radio installation...

46 CFR 91.60-10 - Cargo Ship Safety Equipment Certificate.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Cargo Ship Safety Equipment Certificate. 91.60-10... VESSELS INSPECTION AND CERTIFICATION Certificates Under International Convention for Safety of Life at Sea, 1974 § 91.60-10 Cargo Ship Safety Equipment Certificate. (a) All vessels on an international voyage are...

A Multi-Agent System for Tracking the Intent of Surface Contacts in Ports and Waterways

National Research Council Canada - National Science Library

Tan, Kok S

2005-01-01

...) and employ them to identify asymmetric maritime threats in port and waterways. Each surface track is monitored by a compound multi-agent system that comprise of the several intent models, each containing a nested multi-agent system...

76 FR 77119 - Special Local Regulations; Pompano Beach Holiday Boat Parade, Intracoastal Waterway, Pompano...

Science.gov (United States)

2011-12-12

...-AA08 Special Local Regulations; Pompano Beach Holiday Boat Parade, Intracoastal Waterway, Pompano Beach... Pompano Beach Holiday Boat Parade on Sunday, December 11, 2011. The marine parade will consist of... Friday, except Federal holidays. FOR FURTHER INFORMATION CONTACT: If you have questions on this temporary...

Evaluation of a coolant injection into the in-vessel with a RCS depressurization by using SCDAP/RELAP5

International Nuclear Information System (INIS)

Rae-Joon, Park; Sang-Baik, Kim; Hee-Dong, Kim

2007-01-01

As part of the evaluations of a severe accident management strategy, a coolant injection in the vessel with a reactor coolant system (RCS) depressurization has been evaluated by using the SCDAP/RELAP5 computer code. Two high pressure sequences of a small break loss of coolant accident (LOCA) without safety injection (SI) and a total loss of feed water (LOFW) accident have been analyzed in optimized power reactor OPR-1000. The SCDAP/RELAP5 results have shown that only one train operation of a high pressure safety injection at 30,000 seconds with a RCS depressurization by using one condenser dump valve at 6 minutes after an entrance of the severe accident management guidance prevents a reactor vessel failure for the small break LOCA without SI. In this case, only train operation of the low pressure safety injection (LPSI) without the high pressure safety injection (HPSI) does not prevent a reactor vessel failure. Only one train operation of the HPSI at 20,208 seconds with a RCS depressurization by using two safety depressurization system valves at 40 minutes after an initial opening of the safety relief valve prevents a reactor vessel failure for the total LOFW. (authors)

Safety device for nuclear reactors

International Nuclear Information System (INIS)

Gruhl, H.

1974-01-01

The safety device is used to capture fragments of the lid of a pressure vessel when this vessel ruptures. It consists of a catcher structure attached to the concrete vessel, which is open at the top, and surrounding the pressure vessel. The catcher structure in this case may be designed as a ring installed very close to the concrete vessel, as a closure plate or may be made of transverse beams arranged parallel to each other. It is anchored either rigidly or elastically to the concrete vessel by means of springs or to the foundation by means of steel stretching members. (DG) [de

46 CFR 91.60-15 - Cargo Ship Safety Radio Certificate.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Cargo Ship Safety Radio Certificate. 91.60-15 Section 91... VESSELS INSPECTION AND CERTIFICATION Certificates Under International Convention for Safety of Life at Sea, 1974 § 91.60-15 Cargo Ship Safety Radio Certificate. Every vessel equipped with a radio installation on...

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

International Nuclear Information System (INIS)

Corradini, Michael L.

2006-01-01

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

Structural failure analysis of reactor vessels due to molten core debris

International Nuclear Information System (INIS)

Pfeiffer, P.A.

1993-01-01

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

Assessment and Management of ageing of major nuclear power plant components important to safety: PWR pressure vessels

International Nuclear Information System (INIS)

1999-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g., caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), including water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs; and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. Since the reports are written from a safety perspective, they do not address life or life-cycle management of the plant components, which involves the integration of

F4E R and D programme and results on in-vessel dust and tritium

International Nuclear Information System (INIS)

Le Guern, F.; Gulden, W.; Ciattaglia, S.; Counsell, G.; Bengaouer, A.; Brinster, J.; Dabbene, F.; Denkevitz, A.; Jordan, T.; Kuznetsov, M.; Porfiri, M.T.; Redlinger, R.; Roblin, Ph.; Roth, J.; Segre, J.; Sugiyama, K.; Tkatschenko, I.; Xu, Z.

2011-01-01

In a Tokamak vacuum vessel, plasma-wall interactions can result in the production of radioactive dust and H isotopes (including tritium) can be trapped both in in-vessel material and in dust. The vacuum vessel represents the most important confinement barrier to this radioactive material. In the event of an accident involving ingress of steam to the vacuum vessel, hydrogen could be produced by chemical reactions with hot metal and dust. Hydrogen isotopes could also be desorbed from in-vessel components, e.g. cryopumps. In events where an ingress of air to the vacuum vessel occurs, reaction of the air with hydrogen and/or dust therefore cannot be completely excluded. Due to the radiological risks highlighted by the safety evaluation studies for ITER in normal conditions (e.g. in-vessel maintenance chronic release) and accidental ones (e.g. challenge of vacuum vessel tightness in the event of a hydrogen/dust explosion with air), limitations on the accumulation of dust and tritium in the vacuum vessel are imposed as well as controls over the maximum extent of the quantity of accidental air ingress. ITER IO has defined a strategy for the control of in-vessel dust and tritium inventories below the safety limits based primarily on the measurement and removal of dust and tritium. In this context, this paper will report on the efforts under F4E responsibility to develop a number of the new ITER baseline systems. In particular this paper, after a review of safety constraints and ITER strategy, provides the status of: (1) tasks being launched on diagnostics for in-vessel dust inventory measurement, (2) experiments to enrich the data about the effectiveness of desorption of tritium from Be at 350 o C (divertor baking aiming to release significant amount of tritium trapped in Be co-deposit), (3) on-going R and D programme (experimental and numerical simulation) at FZK, CEA and ENEA on in-vacuum vessel H2 dust explosion.

Using chloride and other ions to trace sewage and road salt in the Illinois Waterway

Science.gov (United States)

Kelly, W.R.; Panno, S.V.; Hackley, Keith C.; Hwang, H.-H.; Martinsek, A.T.; Markus, M.

2010-01-01

Chloride concentrations in waterways of northern USA are increasing at alarming rates and road salt is commonly assumed to be the cause. However, there are additional sources of Cl- in metropolitan areas, such as treated wastewater (TWW) and water conditioning salts, which may be contributing to Cl- loads entering surface waters. In this study, the potential sources of Cl- and Cl- loads in the Illinois River Basin from the Chicago area to the Illinois River's confluence with the Mississippi River were investigated using halide data in stream samples and published Cl- and river discharge data. The investigation showed that road salt runoff and TWW from the Chicago region dominate Cl- loads in the Illinois Waterway, defined as the navigable sections of the Illinois River and two major tributaries in the Chicago region. Treated wastewater discharges at a relatively constant rate throughout the year and is the primary source of Cl- and other elements such as F- and B. Chloride loads are highest in the winter and early spring as a result of road salt runoff which can increase Cl- concentrations by up to several hundred mg/L. Chloride concentrations decrease downstream in the Illinois Waterway due to dilution, but are always elevated relative to tributaries downriver from Chicago. The TWW component is especially noticeable downstream under low discharge conditions during summer and early autumn when surface drainage is at a minimum and agricultural drain tiles are not flowing. Increases in population, urban and residential areas, and roadways in the Chicago area have caused an increase in the flux of Cl- from both road salt and TWW. Chloride concentrations have been increasing in the Illinois Waterway since around 1960 at a rate of about 1 mg/L/a. The increase is largest in the winter months due to road salt runoff. Shallow groundwater Cl- concentrations are also increasing, potentially producing higher base flow concentrations. Projected increases in population and

In-service supervision of a prestressed concrete pressure vessel

International Nuclear Information System (INIS)

Zemann, H.; Mayer, N.; Amberg, C.

1985-01-01

On-line measurements of the physical state of a prestressed concrete pressure vessel and a comparison of the distribution of temperature, strain and stress within the concrete member to the optimized statical predictions and the criterions of layout yield to an efficient and economical method of operating the vessel with a high potential of safety. The requirements of instrumentation and the comparison with static calculations are discussed on the prototype vessel at Seibersdorf Research Center during the phase of construction and prestressing, the phase of the first thermal treatment (stabilization), the pressure tests and under the operating conditions of a high temperature reactor (150 0 C/50 bar). (Author)

In-service supervision of a prestressed concrete pressure vessel

International Nuclear Information System (INIS)

Zemann, H.; Weissbacher, L.; Mayer, N.; Amberge, C.

1985-01-01

On-line measurements of the physical state of a prestressed concrete pressure vessel, and comparison with the design predictions of the distribution of temperature, strain and stress within the concrete member and the criteria of layout, provide an efficient and economical method of operating the vessel with a high potential of safety. The requirements of instrumentation and the comparison with static calculations are discussed with reference to the prototype vessel at Seibersdorf Research Centre during the phase of construction and prestressing, the phase of the first thermal treatment (stabilization), the pressure tests and under the operating conditions of a high temperature reactor (150 0 C, 50 bar). (author)

76 FR 55079 - Recreational Vessel Accident Reporting

Science.gov (United States)

2011-09-06

... operators to make decisions aimed at improving boating safety. This information, described in title 33 Code... Coast Guard long after an accident occurs. Incomplete, inaccurate, or late accident information makes... the recreational vessel owner or operator? If so, how many man-hours are required to collect this...

77 FR 40513 - Safety Zone; Major Motion Picture Filming, Atlantic Intracoastal Waterway; Southport, NC

Science.gov (United States)

2012-07-10

... filming of a movie involving explosions and other dangerous stunts on water. The temporary safety zone is... filming of the movie, a delay in enacting this safety zone would be contrary to public interest. Under 5 U... Planning and Review This rule is not a significant regulatory action under section 3(f) of Executive Order...

Method of detecting water leakage in radioactive waste containing vessel

International Nuclear Information System (INIS)

Ishioka, Hitoshi; Takao, Yoshiaki; Hayakawa, Kiyoshige.

1989-01-01

Lower level radioactive wastes formed upon operation of nuclear facilities are processed by underground storage. In this case, a plurality of drum cans packed with radioactive wastes are contained in a vessel and a water soluble dye material is placed at the inside of the vessel. The method of placing the water soluble dye material at the inside of the vessel includes a method of coating the material on the inner surface of the vessel and a method of mixing the material in sands to be filled between each of the drum cans. Then, leakage of water soluble dye material is detected when water intruding from the outside into the vessel is again leached out of the vessel, to detect the water leakage from the inside of the vessel. In this way, it is possible to find a water-invaded vessel before corrosion of the drum can by water intruded into the vessel and leakage of nuclides in the drum can. Accordingly, it is possible to apply treatment such as repair before occurrence of accident and can maintain the safety of radioactive water processing facilities. (I.S.)

46 CFR 71.75-5 - Passenger Ship Safety Certificate.

Science.gov (United States)

2010-10-01

... 46 Shipping 3 2010-10-01 2010-10-01 false Passenger Ship Safety Certificate. 71.75-5 Section 71.75-5 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PASSENGER VESSELS INSPECTION AND... Passenger Ship Safety Certificate. (a) All vessels on an international voyage are required to have a...

The impact of ornithogenic inputs on phosphorous transport from altered wetland soils to waterways in East Mediterranean ecosystem.

Science.gov (United States)

Litaor, M Iggy; Reichmann, O; Dente, E; Naftaly, A; Shenker, M

2014-03-01

Large flocks of Eurasian crane (Grus grus, >35,000) have begun wintering in an altered wetland agro-ecosystem located in Northern Israel, a phenomenon that attracts more than 400,000 eco-tourists a year. A 100-ha plot has been used to feed the cranes in order to protect nearby fields. The objective of this study was to evaluate the influence of this bird's feeding practice on the P status of the altered wetland soils and waterways. We installed a series of wells at two depths (40 and 90 cm) between two major waterways in the feeding area and monitored the hydraulic heads and collected groundwater samples for elemental analyses. We collected six soil cores and four sediment samples from the waterways and conducted sequential P extraction. We found significant increase in groundwater soluble reactive P (SRP) (>0.5 mg l(-1)) compared with much lower concentrations (~0.06 mg l(-1)) collected in the period prior to the feeding. We found significant decrease in Fe((II)), Ca, and SO4 concentrations in the shallow groundwater (33, 208, and 213 mg l(-1), respectively) compared with the period prior to the feeding (47, 460, and 370 mg l(-1) respectively). An increase in the more labile P fraction was observed in soils and sediments compared with the period before the feeding. The P input by bird excrement to the feeding area was estimated around 700 kg P per season, while P removal by plant harvesting was estimated around 640 kg Pyr(-1). This finding supports the current eco-tourism practices in the middle of intensive farming area, suggesting little impact on waterways. Copyright © 2013 Elsevier B.V. All rights reserved.

Exploring the links between natural resource use and biophysical status in the waterways of the North Rupununi, Guyana.

Science.gov (United States)

Mistry, Jayalaxshmi; Simpson, Matthews; Berardi, Andrea; Sandy, Yung

2004-09-01

The North Rupununi District in south-west Guyana is comprised of a mosaic of ecosystems, including savannas, wetlands and forests, and is home to the Makushi Amerindians, who depend on the waterways for their subsistence needs. With logging and mining seen as increasing threats to the region, it is necessary to look at methods for engaging stakeholders in monitoring the status of their natural resources. This paper presents the results of a pilot study carried out to investigate water use by the Makushi Amerindians, and collect baseline data on the hydro-morphological aspects of the waterways. Methods included informal interviews, the use of the River Habitat Survey (RHS), and water quality measurements. The results indicate the heavy reliance of the Makushi on the waterways for their daily lives, particularly on fishing. Although ponds and creeks are important sources of fish, the rivers provide much larger catches of a greater diversity of fish species, both in the wet and dry seasons. The physical characteristics of the water sources used by the Makushi are mainly associated with the surrounding habitat types: the savanna areas containing the more nutrient rich white-water rivers, and the tropical forest areas containing the less nutrient rich black-water rivers. This study indicates that at present there is no direct evidence of adverse impacts on the waterways used by the Makushi in terms of fish catches, habitat conditions and water quality. A monitoring scheme was set up using this study's outputs as a baseline from which any future changes can be compared. Further work is to be carried out over the next three years to produce monitoring and sustainable management procedures for the North Rupununi ecosystems, by linking the physical attributes of the environment to biodiversity and subsequently local livelihoods, and by building capacity of local stakeholders through training.

Analysis of In-Vessel Late Phase Melt Progression Using SCDAP/RELAP5/MOD3.3

International Nuclear Information System (INIS)

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

2004-01-01

High-pressure in-vessel melt progressions of the KSNP (Korean Standard Nuclear Power Plant) have been analyzed using the SCDAP/RELAP5/MOD3.3 computer code. The total loss of feed water (LOFW) to the steam generators with/without intentional RCS depressurization using the safety depressurization system (SDS) and the station blackout (SBO) have been simulated from transient initiation to reactor vessel failure. The SCDAP/RELAP5/MOD3.3 results have shown that the pressure boundary of the reactor coolant system did not fail before reactor vessel failure in the high-pressure sequences of the LOFW and the SBO transients of the KSNP. In all the high-pressure transients, approximately 20-30 % of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of reactor vessel failure. Intentional RCS depressurization using the SDS for the total LOFW delays reactor vessel failure for approximately 5 hours by actuation of the safety injection tanks. At the time of reactor vessel failure, approximately 50-60 % of the fuel rod cladding was oxidized for the total LOFW and the SBO transients of the KSNP. (authors)

Pressure Safety Program Implementation at ORNL

Energy Technology Data Exchange (ETDEWEB)

Lower, Mark [ORNL; Etheridge, Tom [ORNL; Oland, C. Barry [XCEL Engineering, Inc.

2013-01-01

The Oak Ridge National Laboratory (ORNL) is a US Department of Energy (DOE) facility that is managed by UT-Battelle, LLC. In February 2006, DOE promulgated worker safety and health regulations to govern contractor activities at DOE sites. These regulations, which are provided in 10 CFR 851, Worker Safety and Health Program, establish requirements for worker safety and health program that reduce or prevent occupational injuries, illnesses, and accidental losses by providing DOE contractors and their workers with safe and healthful workplaces at DOE sites. The regulations state that contractors must achieve compliance no later than May 25, 2007. According to 10 CFR 851, Subpart C, Specific Program Requirements, contractors must have a structured approach to their worker safety and health programs that at a minimum includes provisions for pressure safety. In implementing the structured approach for pressure safety, contractors must establish safety policies and procedures to ensure that pressure systems are designed, fabricated, tested, inspected, maintained, repaired, and operated by trained, qualified personnel in accordance with applicable sound engineering principles. In addition, contractors must ensure that all pressure vessels, boilers, air receivers, and supporting piping systems conform to (1) applicable American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (2004) Sections I through XII, including applicable code cases; (2) applicable ASME B31 piping codes; and (3) the strictest applicable state and local codes. When national consensus codes are not applicable because of pressure range, vessel geometry, use of special materials, etc., contractors must implement measures to provide equivalent protection and ensure a level of safety greater than or equal to the level of protection afforded by the ASME or applicable state or local codes. This report documents the work performed to address legacy pressure vessel deficiencies and comply

The need to pressure test prestressed concrete reactor vessels

International Nuclear Information System (INIS)

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

1983-01-01

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

VESSEL-SOURCED POLLUTION: A SECURITY THREAT IN ...

African Journals Online (AJOL)

and some other conventions make provisions concerning protection of ma- ... the pollution of the marine in Malaysia, it appears that pollution by vessels .... pollution from ships and maritime safety; providing effective legal, technical and scientific ..... of the offence after the service of the notice on the offending ship through.

Evaluation of Agency Non-Code Layered Pressure Vessels (LPVs) . Volume 2; Appendices

Science.gov (United States)

Prosser, William H.

2014-01-01

In coordination with the Office of Safety and Mission Assurance and the respective Center Pressure System Managers (PSMs), the NASA Engineering and Safety Center (NESC) was requested to formulate a consensus draft proposal for the development of additional testing and analysis methods to establish the technical validity, and any limitation thereof, for the continued safe operation of facility non-code layered pressure vessels. The PSMs from each NASA Center were asked to participate as part of the assessment team by providing, collecting, and reviewing data regarding current operations of these vessels. This document contains the appendices to the main report.

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)

Research note: Mapping spatial patterns in sewer age, material, and proximity to surface waterways to infer sewer leakage hotspots

Science.gov (United States)

Hopkins, Kristina G.; Bain, Daniel J.

2018-01-01

Identifying areas where deteriorating sewer infrastructure is in close proximity to surface waterways is needed to map likely connections between sewers and streams. We present a method to estimate sewer installation year and deterioration status using historical maps of the sewer network, parcel-scale property assessment data, and pipe material. Areas where streams were likely buried into the sewer system were mapped by intersecting the historical stream network derived from a 10-m resolution digital elevation model with sewer pipe locations. Potential sewer leakage hotspots were mapped by identifying where aging sewer pipes are in close proximity (50-m) to surface waterways. Results from Pittsburgh, Pennsylvania (USA), indicated 41% of the historical stream length was lost or buried and the potential interface between sewers and streams is great. The co-location of aging sewer infrastructure (>75 years old) near stream channels suggests that 42% of existing streams are located in areas with a high potential for sewer leakage if sewer infrastructure fails. Mapping the sewer-stream interface provides an approach to better understand areas were failing sewers may contribute a disproportional amount of nutrients and other pathogens to surface waterways.

Effect of climate change on crop production patterns with implications to transport flows and inland waterways.

Science.gov (United States)

2011-12-01

This project analyzed the demand for transportation capacity and changes in transportation flows on : inland waterways due to shifts in crop production patterns induced by climate change. Shifts in the crop : production mix have been observed in rece...

Stress criteria for nuclear vessel concrete

International Nuclear Information System (INIS)

Costes, D.

1975-01-01

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

Reactor pressure vessel embrittlement

International Nuclear Information System (INIS)

1992-07-01

Within the framework of the IAEA extrabudgetary programme on the Safety of WWER-440/230 NPPs, a list of safety issues requiring broad studies of generic interest have been agreed upon by an Advisory Group who met in Vienna in September 1990. The list was later revised in the light of the programme findings. The information on the status of the issues, and on the amount of work already completed and under way in the various countries, needs to be compiled. Moreover, an evaluation of what further work is required to resolve each one of the issues is also necessary. In view of this, the IAEA has started the preparation of a series of status reports on the various issues. This report on the generic safety issue ''Reactor Pressure Vessel Embrittlement'' presents a comprehensive survey of technical information available in the field and identifies those aspects which require further investigation. 39 refs, 21 figs, 4 tabs

Evaluation of Hydraulic Fracturing (Fracking) Plays for Potential Impact on USACE-Managed Waterways

Science.gov (United States)

2015-01-01

Approved for public release; distribution is unlimited. ERDC/TN DOTS-15-1 January 2015 Evaluation of Hydraulic Fracturing ( Fracking ) Plays...fracturing operations (hydrofracturing or “ fracking ”) to increase petrochemical (natural gas and petroleum) production resulted in elevated environmental...Shale, has raised concerns that fracking operations could impact waterways managed by the U.S. Army Corps of Engineers (USACE). The purpose of this

Structural analysis and evaluation for the design of pressure vessel

International Nuclear Information System (INIS)

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

1977-01-01

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

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

International Nuclear Information System (INIS)

Xiong Dunshi; He Shuyan; Liu Junjie; Yu Suyuan

1993-01-01

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

33 CFR 161.12 - Vessel operating requirements.

Science.gov (United States)

2010-07-01

....0′ N. extending eastward through the Golden Gate, and the navigable waters of San Francisco Bay and... safety beyond that provided by other means. The bridge-to-bridge navigational frequency, 156.650 MHz (Ch... Measures, and Operating Requirements § 161.12 Vessel operating requirements. (a) Subject to the exigencies...

Application of the Streamflow Prediction Tool to Estimate Sediment Dredging Volumes in Texas Coastal Waterways

Science.gov (United States)

Yeates, E.; Dreaper, G.; Afshari, S.; Tavakoly, A. A.

2017-12-01

Over the past six fiscal years, the United States Army Corps of Engineers (USACE) has contracted an average of about a billion dollars per year for navigation channel dredging. To execute these funds effectively, USACE Districts must determine which navigation channels need to be dredged in a given year. Improving this prioritization process results in more efficient waterway maintenance. This study uses the Streamflow Prediction Tool, a runoff routing model based on global weather forecast ensembles, to estimate dredged volumes. This study establishes regional linear relationships between cumulative flow and dredged volumes over a long-term simulation covering 30 years (1985-2015), using drainage area and shoaling parameters. The study framework integrates the National Hydrography Dataset (NHDPlus Dataset) with parameters from the Corps Shoaling Analysis Tool (CSAT) and dredging record data from USACE District records. Results in the test cases of the Houston Ship Channel and the Sabine and Port Arthur Harbor waterways in Texas indicate positive correlation between the simulated streamflows and actual dredging records.

Follow-up Study of ITER Safety Analysis : Large In-vessel First Wall Pipe Break with Wet Confinement Bypass

Energy Technology Data Exchange (ETDEWEB)

Moon, Sung Bo; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-10-15

Previous researches have been analyzed risk assessments of fusion reactors that are dangerous in the severe accidents where the radioactive material released from confinement building to the environment. To simulate the severe accidents in ITER, a number of thermal hydraulics simulation codes were used. Before construction of the fusion reactor, to obtain ITER license about safety issue, MELCOR is chosen as one of the several codes to be used to perform ITER safety analyses. Qualification of the simulation code is to simulate the cooling system in ITER, the transport of radionuclides during design basis accidents (DBAs) including beyond design basis accidents (BDBAs). MELCOR is fully integrated code that models the accidents in Light Water Reactor (LWR). To analyze the accidents in ITER, MELCOR 1.8.2 version is modified. In the nuclear fusion system, the amount of released radioactive material is criteria for safety permission. Tritium (or tritiated water: HTO) and radioactive dust aerosol are the source of radioactive leakage. In the Generic Site Safety Report (GSSR) for the ITER plant, Table I lists the release guidelines for tritium and activation products for normal operation, incidents and accidents. Several accident analyses have been studied to know how much radioactive material could be released from the severe accidents. In the present work, The MELCOR input deck of large First Wall (FW) coolant leak (pipe break) is used to study and radioactive material leakage thorough bypass accident are studied to follow up the ITER safety analysis. In this research, follow-up study of the in-vessel inboard/inboard-outboard FW pipe break was analyzed to investigate the amount of leakage of radioactive aerosol. All of the accident cases released the lower amount of radioactive aerosol compared to the IAEA guide lines. In addition, the OBB pipe break made lower HTO aerosol leakage because of condensation of HTO and adsorption between coolant and aerosol.

The Coast Guard Proceedings of the Marine Safety and Security Council: Spring 2016

Science.gov (United States)

2016-04-01

management system designed to manage safety elements in the workplace . In practice, an operational...winning DuPont family of workplace safety training offerings. Management Buy-In Even if an organization embraces nonconformities as a call to improved...PROCEEDINGS Spring 2016 Vol. 73, Number 1 Safety Management System Objectives 6 Safety Management Facilitates Safe Vessel Operation Vessel

78 FR 75359 - Waterway Suitability Assessment for Construction and Operation of Liquefied Gas Terminals; Orange...

Science.gov (United States)

2013-12-11

... Assessment for Construction and Operation of Liquefied Gas Terminals; Orange, TX AGENCY: Coast Guard, DHS... waterfront facility handling and storing Liquefied Hazardous Gas (LHG) at its Orange, Texas facility. The... LHG marine traffic in the associated waterway. INVISTA, S.a.r.l. located in Orange, Texas submitted an...

PWG4 perspective on ex-vessel hydrogen sources

International Nuclear Information System (INIS)

2000-07-01

The purpose of this perspective document is to identify the potential ex-vessel hydrogen sources and to address the question whether, considered the uncertainties associated to these sources, further investigations are required. The statement is established with reference to the needs for safety evaluation of nuclear reactors under severe accident conditions. It is recognised that the views could be different if one looks at these issues from another standpoint. Since the TMI-2 accident in 1979, there had been a large interest in the nuclear reactor safety community for studying the behaviour of hydrogen in case of a severe accident. As a result, different 'state of the art' reports were produced. Examples of these documents are NUREG/CR-1561 and EUR 14307. In particular, they identified potential hydrogen sources during accidents, including ex-vessel sources. Various ex-vessel hydrogen sources, covering a variety of physical and chemical processes, were identified. Although their precise quantification and relative importance is to be established on a case by case basis with respect to the specific reactor design of interest, general trends can be formulated. The sources to be considered are the followings: - radiolysis of water; - corrosion reactions, - reaction of urania with steam and water; - core-concrete interaction; - debris-atmosphere interaction. These sources are discussed successively. The PWG4 (CSNI's Principal Working Group on the Confinement of Accidental Radioactive Releases) perspective on Ex-vessel Hydrogen Sources can be summarised in the following statements: 1. The issue of hydrogen sources must be considered as a whole and cannot be separated into in-vessel and ex-vessel issues. For significant sources that may not be accommodated by mitigation means associated to DBA, the uncertainty is largely dominated by the unknown extent of Zr oxidation during the in-vessel phase. 2. PWG4 notes that hydrogen production during corium quenching by water is

Aging impact on the safety and operability of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Pennell, W.E.

1992-01-01

Irradiation embrittlement causes a loss of reactor vessel material fracture toughness as nuclear plants age. Fracture mechanics based regulatory requirements limit the permissible level of irradiation embrittlement such that essential fracture prevention margins are maintained throughout the plant operating life. This paper reviews the regulatory requirements and the underlying fracture mechanics technology. Issues identified with that technology are identified and research programs implemented to resolve the issues are described. Where possible, an assessment is given of the anticipated impact on the research program output will have on the reactor vessel fracture-margin assessment process

Seismic proving test of PWR reactor containment vessel

International Nuclear Information System (INIS)

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

1987-01-01

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

Assessment and management of ageing of major nuclear power plant components important to safety: PWR pressure vessels. 2007 update

International Nuclear Information System (INIS)

2007-06-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that effective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) is one of the most important issues for plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. IAEA-TECDOC-1120 documented ageing assessment and management practices for pressurized water reactor (PWR) reactor pressure vessels (RPVs) that were current at the time of its finalization in 1997-1998. Safety significant operating events have occurred since the finalization of the TECDOC, e.g. primary water stress corrosion cracking (PWSCC) of Alloy 600 control rod drive mechanism (CRDM) penetrations and boric acid corrosion/wastage of RPV heads, which threatened the integrity of the RPV heads. These events led to new ageing management actions by both NPP operators and regulators. Therefore it was recognized that IAEA-TECDOC-1120 should be updated by incorporating those new events and their countermeasures. The objective of this report is to update IAEA-TECDOC-1120 in order to provide current ageing management guidance for PWR RPVs to all involved in the operation and regulation of PWRs and thus to help ensure PWR RPV integrity in IAEA Member States throughout their entire service life

Provision of reliable core cooling in vessel-type boiling reactors

International Nuclear Information System (INIS)

Alferov, N.S.; Balunov, B.F.; Davydov, S.A.

1987-01-01

Methods for providing reliable core cooling in vessel-type boiling reactors with natural circulation for heat supply are analysed. The solution of this problem is reduced to satisfaction of two conditions such as: water confinement over the reactor core necessary in case of an accident and confinement of sufficient coolant flow rate through the bottom cross section of fuel assemblies for some time. The reliable fuel element cooling under conditions of a maximum credible accident (brittle failure of a reactor vessel) is shown to be provided practically in any accident, using the safety vessel in combination with the application of means of standard operation and minimal composition and capacity of ECCS

In-place thermal annealing of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Server, W.L.

1985-04-01

Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. The Amry SM-1A test reactor vessel was wet annealed in 1967 at less than 343 0 C (650 0 F), and wet annealing of the Belgian BR-3 reactor vessel at 343 0 C (650 0 F) has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place at temperatures as high as 454 0 C (850 0 F) is feasible, but solvable engineering problems do exist. Economic considerations have not been totally evaluated in assessing the cost-effectiveness of in-place annealing of commercial nuclear vessels. An American Society for Testing and Materials (ASTM) task group is upgrading and revising guide ASTM E 509-74 with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (e.g., the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

Performance Analysis of Multi Stage Safety Injection Tank

International Nuclear Information System (INIS)

Shin, Soo Jai; Kim, Young In; Bae, Youngmin; Kang, Han-Ok; Kim, Keung Koo

2015-01-01

In general the integral reactor has such characteristics, the integral reactor requires a high flow rate of coolant safety injection at the initial stage of the accident in which the core level is relatively fast decreased, A medium flow rate of coolant safety injection at the early and middle stages of the accident in which the coolant discharge flow rate is relatively large due to a high internal pressure of the reactor vessel, and a low flow rate of coolant safety injection is required at the middle and late stages of the accident in which the coolant discharge flow rate is greatly reduced due to a decreased pressure of the reactor vessel. It is noted that a high flow rate of the integral reactor is quite smaller compared to a flow rate required in the commercial loop type reactor. However, a nitrogen pressurized safety injection tank has been typically designed to quickly inject a high flow rate of coolant when the internal pressure of the reactor vessel is rapidly decreased, and a core makeup tank has been designed to safely inject at a single mode flow rate due to a gravitational head of water subsequent to making a pressure balance between the reactor vessel and core makeup tank. As a result, in order to compensate such a disadvantage, various type systems are used in a complicated manner in a reactor according to the required characteristic of safety injection during an accident. In the present study, we have investigated numerically the performance of the multi stage safety injection tank. A parameter study has performed to understand the characteristics of the multi stage safety injection tank. The performance of the multi stage safety injection tank has been investigated numerically. When an accident occurs, the coolant in the multi stage safety injection tank is injected into a reactor vessel by a gravitational head of water subsequent to making a pressure balance between the reactor and tank. At the early stages of the accident, the high flow rate of

Performance Analysis of Multi Stage Safety Injection Tank

Energy Technology Data Exchange (ETDEWEB)

Shin, Soo Jai; Kim, Young In; Bae, Youngmin; Kang, Han-Ok; Kim, Keung Koo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In general the integral reactor has such characteristics, the integral reactor requires a high flow rate of coolant safety injection at the initial stage of the accident in which the core level is relatively fast decreased, A medium flow rate of coolant safety injection at the early and middle stages of the accident in which the coolant discharge flow rate is relatively large due to a high internal pressure of the reactor vessel, and a low flow rate of coolant safety injection is required at the middle and late stages of the accident in which the coolant discharge flow rate is greatly reduced due to a decreased pressure of the reactor vessel. It is noted that a high flow rate of the integral reactor is quite smaller compared to a flow rate required in the commercial loop type reactor. However, a nitrogen pressurized safety injection tank has been typically designed to quickly inject a high flow rate of coolant when the internal pressure of the reactor vessel is rapidly decreased, and a core makeup tank has been designed to safely inject at a single mode flow rate due to a gravitational head of water subsequent to making a pressure balance between the reactor vessel and core makeup tank. As a result, in order to compensate such a disadvantage, various type systems are used in a complicated manner in a reactor according to the required characteristic of safety injection during an accident. In the present study, we have investigated numerically the performance of the multi stage safety injection tank. A parameter study has performed to understand the characteristics of the multi stage safety injection tank. The performance of the multi stage safety injection tank has been investigated numerically. When an accident occurs, the coolant in the multi stage safety injection tank is injected into a reactor vessel by a gravitational head of water subsequent to making a pressure balance between the reactor and tank. At the early stages of the accident, the high flow rate of

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.

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

International Nuclear Information System (INIS)

Chen Liying; Guo Jilin; Liu Wei

2000-01-01

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

RB research reactor safety report

International Nuclear Information System (INIS)

Sotic, O.; Pesic, M.; Vranic, S.

1979-04-01

This new version of the safety report is a revision of the safety report written in 1962 when the RB reactor started operation after reconstruction. The new safety report was needed because reactor systems and components have been improved and the administrative procedures were changed. the most important improvements and changes were concerned with the use of highly enriched fuel (80% enriched), construction of reactor converter outside the reactor vessel, improved control system by two measuring start-up channels, construction of system for heavy water leak detection, new inter phone connection between control room and other reactor rooms. This report includes description of reactor building with installations, rector vessel, reactor core, heavy water system, control system, safety system, dosimetry and alarm systems, experimental channels, neutron converter, reactor operation. Safety aspects contain analyses of accident reasons, method for preventing reactivity insertions, analyses of maximum hypothetical accidents for cores with natural uranium, 2% enriched and 80% enriched fuel elements. Influence of seismic events on the reactor safety and well as coupling between reactor and the converter are parts of this document

Research to sustain cases for Magnox-reactor steel pressure vessels

International Nuclear Information System (INIS)

Graham, W.J.

1997-01-01

Britain's Magnox Electric plc owns and operates six power stations, each of which has twin gas-cooled reactors of the Magnox-fuel type. The older group of four power stations has steel pressure-circuits. The reactor cores are housed within spherical, steel vessels. This article describes some of the research which is undertaken to sustain the safety cases for these steel vessels which have now been in operation for just over 30 years. (author) 2 figs., 4 refs

26 CFR 48.4042-2 - Special rules.

Science.gov (United States)

2010-04-01

... TAXES MANUFACTURERS AND RETAILERS EXCISE TAXES Fuel Used on Inland Waterways § 48.4042-2 Special rules...)); (ii) The type of vessel in which fuel is consumed and the type of vessel in which cargo is transported...

Proposal of In-vessel corium retention concept for Paks NPP

International Nuclear Information System (INIS)

Elter, J.; Toth, E.; Matejovic, P.

2011-01-01

The in-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) seems to be a promising severe accident management strategy not only for new generation of advanced PWRs, but also for VVER-440/V213 reactors, which were designed several years ago. The basic idea of in-vessel retention of corium is to prevent RPV failure by flooding the reactor cavity so that the reactor pressure vessel is submerged in water up to its support structures, and thus the decay heat can be transferred from the corium pool through the vessel wall and into the water surrounding the vessel. An IVR concept with simple ECVR loop based only on minor modifications of existing plant technology was proposed for the Paks Nuclear Power Plant. 2 severe accident (LB and SB LOCA) without availability of HP and LP safety injection in power upgrade (108%) conditions were simulated using the ASTEC code. The analyses show that the proposed solution is effective in preserving RPV integrity in the case of severe accident. Possible uncertainties in code predictions are covered by the applied conservative assumptions

76 FR 78151 - Special Local Regulations; Boca Raton Holiday Boat Parade, Intracoastal Waterway, Boca Raton, FL

Science.gov (United States)

2011-12-16

...-AA08 Special Local Regulations; Boca Raton Holiday Boat Parade, Intracoastal Waterway, Boca Raton, FL... Raton Holiday Boat Parade on Saturday, December 17, 2011. The marine parade will consist of... through Friday, except Federal holidays. FOR FURTHER INFORMATION CONTACT: If you have questions on this...

75 FR 6215 - Houston/Galveston Navigation Safety Advisory Committee

Science.gov (United States)

2010-02-08

... groups will meet in Houston, Texas to discuss waterway improvements, aids to navigation, area projects...; (b) Dredging subcommittee report; (c) Technology subcommittee report; (d) Waterways Optimization...

77 FR 39406 - Safety Zone; Tom Graves Memorial Fireworks, Port Bay, Wolcott, NY

Science.gov (United States)

2012-07-03

...-AA00 Safety Zone; Tom Graves Memorial Fireworks, Port Bay, Wolcott, NY AGENCY: Coast Guard, DHS. ACTION..., NY. This safety zone is intended to restrict vessels from a portion of Port Bay during the Tom Graves... necessary to ensure the safety of spectators and vessels during the Tom Graves Memorial Fireworks. This zone...

Soot concentrations along busy inland waterways in the Netherlands; Roetconcentraties langs drukke binnenvaarwegen in Nederland

Energy Technology Data Exchange (ETDEWEB)

Keuken, M.; Jonkers, S.; Moerman, M. [TNO Built Environment and Geosciences, Delft (Netherlands); Hoek, G. [Institute for Risk Assessment Sciences IRAS, Utrecht University, Utrecht (Netherlands)

2013-08-15

Exploratory research by TNO and IRAS shows that residents in the Netherlands are exposed to soot concentrations along busy inland waterways similar to living along a busy highway [Dutch] Verkennend onderzoek van TNO en IRAS laat zien dat bewoners langs drukke binnenvaarwegen worden blootgesteld aan roetconcentraties vergelijkbaar met wonen langs een drukke snelweg.

AMNT 2014. Key topic: Reactor operation, safety - report. Pt. 2

International Nuclear Information System (INIS)

Fischer, Klaus-Christian; Willschuetz, Hans-Georg; Wortmann, Birgit

2014-01-01

Summary report on the following sessions of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - Thermo Dynamics and Fluid Dynamics: Experiments and Backfittings for the Improvement of Safety and Efficiency; - Safety of Nuclear Installations - Methods, Analyses, Results: In-Vessel Phenomena; Ex-Vessel Phenomena; - Standards and Regulations; Hazard and Safety Analysis; and Validation and Uncertainty Analysis. The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' have been covered in atw 10 (2014) and will be covered in further issues of atw.

Cyanobacteria of the 2016 Lake Okeechobee and Okeechobee Waterway harmful algal bloom

Science.gov (United States)

Rosen, Barry H.; Davis, Timothy W.; Gobler, Christopher J.; Kramer, Benjamin J.; Loftin, Keith A.

2017-05-31

The Lake Okeechobee and the Okeechobee Waterway (Lake Okeechobee, the St. Lucie Canal and River, and the Caloosahatchee River) experienced an extensive harmful algal bloom within Lake Okeechobee, the St. Lucie Canal and River and the Caloosahatchee River in 2016. In addition to the very visible bloom of the cyanobacterium Microcystis aeruginosa, several other cyanobacteria were present. These other species were less conspicuous; however, they have the potential to produce a variety of cyanotoxins, including anatoxins, cylindrospermopsins, and saxitoxins, in addition to the microcystins commonly associated with Microcystis. Some of these species were found before, during, and 2 weeks after the large Microcystis bloom and could provide a better understanding of bloom dynamics and succession. This report provides photographic documentation and taxonomic assessment of the cyanobacteria present from Lake Okeechobee and the Caloosahatchee River and St. Lucie Canal, with samples collected June 1st from the Caloosahatchee River and Lake Okeechobee and in July from the St. Lucie Canal. The majority of the images were of live organisms, allowing their natural complement of pigmentation to be captured. The report provides a digital image-based taxonomic record of the Lake Okeechobee and the Okeechobee Waterway microscopic flora. It is anticipated that these images will facilitate current and future studies on this system, such as understanding the timing of cyanobacteria blooms and their potential toxin production.

MARS vessel safety analysis. LATA report No. 115

International Nuclear Information System (INIS)

Rigdon, L.D.; Donham, B.J.; Hughes, P.S.

1979-08-01

A previous study was performed to assess the hazards associated with an accidental leakage of cooling water into the crucible of molten 238 U for the MARS laser isotope separation experiment. Since that study found that the probability of such an explosion is extremely low during an accidental cooling system failure, a study was conducted to define a more realistic design basis accident (DBA) for the final MARS configuration. If the vapor-phase explosion is considered to be a significant threat, the design criteria for the vacuum vessel should be a working pressure of 67 psig or 101 psig momentary single pulse equivalent static pressure

46 CFR 169.723 - Safety belts.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Safety belts. 169.723 Section 169.723 Shipping COAST... Control, Miscellaneous Systems, and Equipment § 169.723 Safety belts. Each vessel must carry a harness type safety belt conforming to Offshore Racing Council (ORC) standards for each person on watch or...

Towards a new pressure vessel standard in the European Union

International Nuclear Information System (INIS)

Osweiller, F.

1995-01-01

Since 1990 the European Commission has been preparing a new Directive which will regulate the Pressure Equipment sector in the countries of the European Union. CEN Standards devoted to pressure vessels, piping, boilers, are currently being drawn up to complete and implement this Directive. This paper focuses on the European Unfired Pressure Vessel Standard (EPVS) which is in course of development under the responsibility of CEN/TC54. The main aspects of the Standard are outlined: general structure, materials, design, fabrication, inspection and testing. The link with the European Directive is explained in connection with regulatory aspects: conformity assessment, essential safety requirements, classes of vessels, notified bodies, EC mark, status of the standard

Reliability aspects of radiation damage in reactor pressure vessel mterials

International Nuclear Information System (INIS)

Brumovsky, M.

1985-01-01

The service life estimate is a major factor in the evaluation of the operating reliability and safety of a nuclear reactor pressure vessel. The evaluation of the service life of the pressure vessel is based on a comparison of fracture toughness values with stress intensity factors. Notch toughness curves are used for the indirect determination of fracture toughness. The dominant degradation effect is radiation embrittlement. Factors having the greatest effect on the result are the properties of the starting material of the vessel and the impurity content, mainly the Cu and P content. The design life is affected by the evaluation of residual lifetime which is made by periodical nondestructive inspections and using surveillance samples. (M.D.)

Role of radiation embrittlement in reactor vessel integrity assessment

International Nuclear Information System (INIS)

Marston, T.U.; Chexal, V.K.; Wyckoff, M.

1982-01-01

Reactor vessel integrity calculations are complex. The effect of radiation embrittlement on vessel material properties is a very important aspect of any vessel integrity evaluation. The importance of realistic (based on surveillance capsule results) rather than conservative estimates of the material properties (based on regulatory curves) cannot be overestimated. It is also important to make realistic thermal hydraulic and system operations assumptions. In addition, use of actual flaw sizes from in-service inspections (versus hypothetical flaw size selection) will promote realism. Important research results exist that need to be incorporated into the regulatory process. The authors believe results from current research and development efforts will demonstrate that, with reasonable assumptions and best estimate calculations, the safety of even the older reactor vessels with high copper content welds can be assured over their design lifetimes without the need for major fixes. The utilities, through EPRI and the vendors, have dedicated a significant effort to solving the pressurized thermal shock problem

Analysis code for pressure in reactor containment vessel of ATR. CONPOL

International Nuclear Information System (INIS)

1997-08-01

For the evaluation of the pressure and temperature in containment vessels in the events which are classified in the abnormal change of pressure, atmosphere and others in reactor containment vessels in accident among the safety evaluation events of the ATR, the analysis code for the pressure in reactor containment vessels CONPOL is used. In this report, the functions of the analysis code and the analysis model are shown. By using this analysis code, the rise of the pressure and temperature in a containment vessel is evaluated when loss of coolant accident occurs, and high temperature, high pressure coolant flows into it. This code possesses the functions of computing blow-down quantity and heat dissipation from reactor cooling facility, steam condensing heat transfer to containment vessel walls, and the cooling effect by containment vessel spray system. As for the analysis techniques, the models of reactor cooling system, containment vessel and steam discharge pool, and the computation models for the pressure and temperature in containment vessels, wall surface temperature, condensing heat transfer, spray condensation and blow-down are explained. The experimental analysis of the evaluation of the pressure and temperature in containment vessels at the time of loss of coolant accident is reported. (K.I.)

In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

Energy Technology Data Exchange (ETDEWEB)

F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

2005-01-01

In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

The 1500 MW fast breeder reactor the double envelope-vessel anchored in concrete

International Nuclear Information System (INIS)

Bolvin, M.

1981-01-01

This paper givers an account of EDF investigations to reduce the investment costs of the 1500 MW Fast Reactor (RNR 1500) without prejudice to the safety requirements. It deals with the double envelope-vessel, designed to minimize radiation consequences in the event of accidental leakage in the main vessel. In the Fast Reactors in operation (PHOENIX), under construction (CRYS-MALVILLE), and under project (NR 1500), the double envelope-steel vessel hangs down from the upper part of the reactor block, its weight being approximately 300 t. In the new design, the vessel is fixed into the concrete which supports the main vessel, by means of steel anchors. A thermal insulation isolates it from the main vessel. The installation of coils in the concrete, next to the lining, allows for water circulation to cool the concrete. (orig./GL)

Dust Combustion Safety Issues for Fusion Applications

Energy Technology Data Exchange (ETDEWEB)

L. C. Cadwallader

2003-05-01

This report summarizes the results of a safety research task to identify the safety issues and phenomenology of metallic dust fires and explosions that are postulated for fusion experiments. There are a variety of metal dusts that are created by plasma erosion and disruptions within the plasma chamber, as well as normal industrial dusts generated in the more conventional equipment in the balance of plant. For fusion, in-vessel dusts are generally mixtures of several elements; that is, the constituent elements in alloys and the variety of elements used for in-vessel materials. For example, in-vessel dust could be composed of beryllium from a first wall coating, tungsten from a divertor plate, copper from a plasma heating antenna or diagnostic, and perhaps some iron and chromium from the steel vessel wall or titanium and vanadium from the vessel wall. Each of these elements has its own unique combustion characteristics, and mixtures of elements must be evaluated for the mixture’s combustion properties. Issues of particle size, dust temperature, and presence of other combustible materials (i.e., deuterium and tritium) also affect combustion in air. Combustion in other gases has also been investigated to determine if there are safety concerns with “inert” atmospheres, such as nitrogen. Several coolants have also been reviewed to determine if coolant breach into the plasma chamber would enhance the combustion threat; for example, in-vessel steam from a water coolant breach will react with metal dust. The results of this review are presented here.

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

International Nuclear Information System (INIS)

Sun Yunxue; Shao Jin

2015-01-01

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

Pressure vessels for reactors made from structural steel with limited tensile strength

International Nuclear Information System (INIS)

Machatti, H.

1973-01-01

The reactor pressure vessel is prestressed in several directions with prestressing elements fabricated of steel with a high yielding point. This design allows a substantial reduction of wall thickness or an increase of the inner diameter at equal wall thickness. The prestress of the prestressing elements is designed to achieve a maximum stress release of the vessel walls at normal operating conditions and to fully utilize the maximum load of the vessel walls. For safety reasons the cross section of the prestressing elements is constructed in a way that strain is always 20 % lower the yield point. (P.K.)

Elimination of the risk of brittle fracture in thick welded pressure vessels

International Nuclear Information System (INIS)

Leymonie, C.; Genevray, R.

1975-01-01

The builder of welded pressure vessels faces the risk of brittle fracture throughout fabrication. He is forced to observe many precautions, in selecting the following: materials possessing good impact strength in the service conditions of the vessels; filler materials preventing transverse cracking of the welds: welding parameters preventing cold cracking. Fracture mechanics establish the relationships between material characteristics and critical defect size for a given set of service conditions. These principles must be expanded to increase the safety of thick pressure vessels. However, in order to derive maximum benefit, a major effort must be applied to increasing the effectiveness of nondestructive testing [fr

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

International Nuclear Information System (INIS)

Stahlberg, R.

1978-01-01

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

Preliminary structural assessment of DEMO vacuum vessel against a vertical displacement event

International Nuclear Information System (INIS)

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

2016-01-01

Highlights: • The paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel. • The Vacuum Vessel was checked against the VDE in combinations with the weight force of all components that the vessel shall bear. • Different configurations for the vacuum vessel supports are considered, showing that the best solution is VV supported at the lower port. • The analyses evaluated the “P damage” according to RCC-MRx code. - Abstract: This paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel (VV). The VV structure is checked against a vertical load due to a Vertical Displacement Event in combination with the weight force of all components that the main vessel shall bear. Different configurations for the supports are considered. Results show that the greatest safety margins are reached when the tokamak is supported through the lower ports rather than the equatorial ports, though all analyzed configurations are compliant with RCC-MRx design rules.

Preliminary structural assessment of DEMO vacuum vessel against a vertical displacement event

Energy Technology Data Exchange (ETDEWEB)

Mozzillo, Rocco, E-mail: rocco.mozzillo@unina.it [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Tarallo, Andrea; Marzullo, Domenico [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Bachmann, Christian [EUROfusion PMU, Boltzmannstraße 2, 85748 Garching (Germany); Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Mazzone, Giuseppe [Unità Tecnica Fusione - ENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy)

2016-11-15

Highlights: • The paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel. • The Vacuum Vessel was checked against the VDE in combinations with the weight force of all components that the vessel shall bear. • Different configurations for the vacuum vessel supports are considered, showing that the best solution is VV supported at the lower port. • The analyses evaluated the “P damage” according to RCC-MRx code. - Abstract: This paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel (VV). The VV structure is checked against a vertical load due to a Vertical Displacement Event in combination with the weight force of all components that the main vessel shall bear. Different configurations for the supports are considered. Results show that the greatest safety margins are reached when the tokamak is supported through the lower ports rather than the equatorial ports, though all analyzed configurations are compliant with RCC-MRx design rules.

Design optimization of a thin walled pressure vessel

International Nuclear Information System (INIS)

Sadiq, S.

2001-01-01

Design evaluation of a pressure vessel is not only to build confidence on its integrity but also to reduce structural weight and enhance the performance of the structure. Pressure vessel, e.g., a rocket motor not only has to withstand the high operating temperatures but it must also be able to survive the internal pressures and external aerodynamic forces and bending stresses during its operation in flight. A research program was devised to study the stresses, which are generated in a thin walled pressure vessel during actual operation and its simulation with cold testing technique, i.e., by means of hydrostatic testing employing electrical resistance strain gauges on the external surface of the cylinder. The objective of the research was to uphold the performance of the vessel by reducing its thickness from 6.09 to 5.5 mm (which of course reduces the safety factor margin from 1.8 to 1.5); thereby curtailing the overall structural weight and maintaining the efficiency of the vessel itself during its live operation. The techniques employed were hydrostatic testing, data acquisition system for obtaining data on strains from the electrical resistance strain gauges and later employing V on Mises yield criterion empirical relation to computer the stresses in hoop and longitudinal directions. (author)

Consequence evaluation of radiation embrittlement of Trojan reactor pressure vessel supports

International Nuclear Information System (INIS)

Lu, S.C.; Sommer, S.C.; Johnson, G.L.; Lambert, H.E.

1990-10-01

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

Total system for manufacture of nuclear vessels by computer: VECTRON

International Nuclear Information System (INIS)

Inagawa, Jin; Ueno, Osamu; Hanai, Yoshiharu; Ohkawa, Isao; Washizu, Hideyuki

1980-01-01

VECTRON (Vessel Engineering by Computer Tool and Rapid Operating for the N/C System) is a CAM (Computer Aided Manufacturing) system that has been developed to produce high quality and highly accurate vessels for nuclear power plants and other industrial plants. Outputs of this system are design drawings, manufacturing information and magnetic tapes of the N/C marking machine for vessel shell plates including their attachments. And it can also output information at each stage of designing, marking, cutting, forming and assembling by treating the vessels in three dimensions and by using data filing systems and plotting program for general use. The data filing systems consist of functional and manufacturing data of each part of vessels. This system not only realizes a change from manual work to computer work, but also leads us to improve production engineering and production jigs for safety and high quality. At present, VECTRON is being applied to the manufacture of the shell plates of primary containment vessels in the Kashiwazaki-Kariwa Nuclear Power Station Unit 1 (K-1) and the Fukushima Daini Nuclear Power Station Unit 3 (2F-3), to realize increased productivity. (author)

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

Design of integrated passive safety system (IPSS) for ultimate passive safety of nuclear power plants

International Nuclear Information System (INIS)

Chang, Soon Heung; Kim, Sang Ho; Choi, Jae Young

2013-01-01

Highlights: • We newly propose the design concept of integrated passive safety system (IPSS). • It has five safety functions for decay heat removal and severe accident mitigation. • Simulations for IPSS show that core melt does not occur in accidents with SBO. • IPSS can achieve the passive in-vessel retention and ex-vessel cooling strategy. • The applicability of IPSS is high due to the installation outside the containment. -- Abstract: The design concept of integrated passive safety system (IPSS) which can perform various passive safety functions is proposed in this paper. It has the various functions of passive decay heat removal system, passive safety injection system, passive containment cooling system, passive in-vessel retention and cavity flooding system, and filtered venting system with containment pressure control. The objectives of this paper are to propose the conceptual design of an IPSS and to estimate the design characters of the IPSS with accident simulations using MARS code. Some functions of the IPSS are newly proposed and the other functions are reviewed with the integration of the functions. Consequently, all of the functions are modified and integrated for simplicity of the design in preparation for beyond design based accidents (BDBAs) focused on a station black out (SBO). The simulation results with the IPSS show that the decay heat can be sufficiently removed in accidents that occur with a SBO. Also, the molten core can be retained in a vessel via the passive in-vessel retention strategy of the IPSS. The actual application potential of the IPSS is high, as numerous strong design characters are evaluated. The installation of the IPSS into the original design of a nuclear power plant requires minimal design change using the current penetrations of the containment. The functions are integrated in one or two large tanks outside the containment. Furthermore, the operation time of the IPSS can be increased by refilling coolant from the

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

Nuclear Reactor RA Safety Report, Vol. 14, Safety protection measures

International Nuclear Information System (INIS)

1986-11-01

Nuclear reactor accidents can be caused by three type of errors: failure of reactor components including (1) control and measuring instrumentation, (2) errors in operation procedure, (3) natural disasters. Safety during reactor operation are secured during its design and construction and later during operation. Both construction and administrative procedures are applied to attain safe operation. Technical safety features include fission product barriers, fuel elements cladding, primary reactor components (reactor vessel, primary cooling pipes, heat exchanger in the pump), reactor building. Safety system is the system for safe reactor shutdown and auxiliary safety system. RA reactor operating regulations and instructions are administrative acts applied to avoid possible human error caused accidents [sr

Influences of the separation distance, ship speed and channel dimension on ship maneuverability in a confined waterway

Science.gov (United States)

Du, Peng; Ouahsine, Abdellatif; Sergent, Philippe

2018-05-01

Ship maneuvering in the confined inland waterway is investigated using the system-based method, where a nonlinear transient hydrodynamic model is adopted and confinement models are implemented to account for the influence of the channel bank and bottom. The maneuvering model is validated using the turning circle test, and the confinement model is validated using the experimental data. The separation distance, ship speed, and channel width are then varied to investigate their influences on ship maneuverability. With smaller separation distances and higher speeds near the bank, the ship's trajectory deviates more from the original course and the bow is repelled with a larger yaw angle, which increase the difficulty of maneuvering. Smaller channel widths induce higher advancing resistances on the ship. The minimum distance to the bank are extracted and studied. It is suggested to navigate the ship in the middle of the channel and with a reasonable speed in the restricted waterway.

ARIES-AT safety design and analysis

Energy Technology Data Exchange (ETDEWEB)

Petti, D.A. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States)]. E-mail: David.Petti@inl.gov; Merrill, B.J. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Moore, R.L. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Longhurst, G.R. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); El-Guebaly, L. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Mogahed, E. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Henderson, D. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Wilson, P. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Abdou, A. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States)

2006-01-15

ARIES-AT is a 1000 MWe conceptual fusion power plant design with a very low projected cost of electricity. The design contains many innovative features to improve both the physics and engineering performance of the system. From the safety and environmental perspective, there is greater depth to the overall analysis than in past ARIES studies. For ARIES-AT, the overall spectrum of off-normal events to be examined has been broadened. They include conventional loss of coolant and loss of flow events, an ex-vessel loss of coolant, and in-vessel off-normal events that mobilize in-vessel inventories (e.g., tritium and tokamak dust) and bypass primary confinement such as a loss of vacuum and an in-vessel loss of coolant with bypass. This broader examination of accidents improves the robustness of the design from the safety perspective and gives additional confidence that the facility can meet the no-evacuation requirement under average weather conditions. We also provide a systematic assessment of the design to address key safety functions such as confinement, decay heat removal, and chemical energy control. In the area of waste management, both the volume of the component and its hazard are used to classify the waste. In comparison to previous ARIES designs, the overall waste volume is less because of the compact design.

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Stress analysis of R2 pressure vessel. Structural reliability benchmark exercise

International Nuclear Information System (INIS)

Vestergaard, N.

1987-05-01

The Structural Reliability Benchmark Exercise (SRBE) is sponsored by the EEC as part of the Reactor Safety Programme. The objectives of the SRBE are to evaluate and improve 1) inspection procedures, which use non-destructive methods to locate defects in pressure (reactor) vessels, as well as 2) analytical damage accumulation models, which predict the time to failure of vessels containing defects. In order to focus attention, an experimental presure vessel has been inspected, subjected fatigue loadings and subsequently analysed by several teams using methods of their choice. The present report contains the first part of the analytical damage accumulation analysis. The stress distributions in the welds of the experimental pressure vessel were determined. These stress distributions will be used to determine the driving forces of the damage accumulation models, which will be addressed in a future report. (author)

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

60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

International Nuclear Information System (INIS)

Duysen, J.C. van; Meric de Bellefon, G.

2017-01-01

The first light water nuclear reactor dedicated to electricity production was commissioned in Shippingport, Pennsylvania in the United States in 1957. Sixty years after the event, it is clear that this type of reactor will be a major source of electricity and one of the key solutions to limit climate change in the 21st century. This article pays homage to the teams that contributed to this achievement by their involvement in research and development and their determination to push back the frontiers of knowledge. Via a few examples of scientific or technological milestones, it describes the evolution of ideas, models, and techniques during the last 60 years, and gives the current state-of-the-art in areas related to the safety of the reactor pressure vessel. Among other topics, it focuses on vessel manufacturing, steel fracture mechanics analysis, and understanding of irradiation-induced damage.

60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Duysen, J.C. van, E-mail: jean-claude.van-duysen@ensc-lille.fr [Department of Nuclear Engineering University of Tennessee Knoxville (United States); Unité Matériaux et Transformation (UMET) CNRS, Université de Lille 1 (France); Meric de Bellefon, G., E-mail: mericdebelle@wisc.edu [Department of Nuclear Engineering, University of Wisconsin, Madison (United States)

2017-02-15

The first light water nuclear reactor dedicated to electricity production was commissioned in Shippingport, Pennsylvania in the United States in 1957. Sixty years after the event, it is clear that this type of reactor will be a major source of electricity and one of the key solutions to limit climate change in the 21st century. This article pays homage to the teams that contributed to this achievement by their involvement in research and development and their determination to push back the frontiers of knowledge. Via a few examples of scientific or technological milestones, it describes the evolution of ideas, models, and techniques during the last 60 years, and gives the current state-of-the-art in areas related to the safety of the reactor pressure vessel. Among other topics, it focuses on vessel manufacturing, steel fracture mechanics analysis, and understanding of irradiation-induced damage.

Seismic safety margin assessment program (Annual safety research report, JFY 2010)

International Nuclear Information System (INIS)

Suzuki, Kenichi; Iijima, Toru; Inagaki, Masakatsu; Taoka, Hideto; Hidaka, Shinjiro

2011-01-01

Seismic capacity test data, analysis method and evaluation code provided by Seismic Safety Margin Assessment Program have been utilized for the support of seismic back-check evaluation of existing plants. The summary of the program in 2010 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. Many seismic capacity tests of various snubbers were conducted and quantitative seismic capacities were evaluated. One of the emergency diesel generator partial-model seismic capacity tests was conducted and quantitative seismic capacity was evaluated. Some of the analytical evaluations of piping-system seismic capacities were conducted. 2. Analysis method for minute evaluation of component seismic response. The difference of seismic response of large components such as primary containment vessel and reactor pressure vessel when they were coupled with 3-dimensional FEM building model or 1-dimensional lumped mass building model, was quantitatively evaluated. 3. Evaluation code for quantitative evaluation of seismic safety margin of systems, structures and components. As the example, quantitative evaluation of seismic safety margin of systems, structures and components were conducted for the reference plant. (author)

33 CFR 165.155 - Northville Industries Offshore Platform, Riverhead, Long Island, New York-safety zone.

Science.gov (United States)

2010-07-01

... York, 1 mile North of the Riverhead shoreline at 41°00″ N, 072°38″ W, while a Liquefied Petroleum Gas (LPG) vessel is moored at the Offshore Platform. The safety zone remains in effect until the LPG vessel... Offshore Platform of LPG vessels via Marine Safety Information Radio Broadcast. [CGD3 85-86, 51 FR 37181...

Widespread infilling of tidal channels and navigable waterways in human-modified tidal deltaplain of southwest Bangladesh

Directory of Open Access Journals (Sweden)

Carol Wilson

2017-12-01

Full Text Available Since the 1960s, ~5000 km2 of tidal deltaplain in southwest Bangladesh has been embanked and converted to densely inhabited, agricultural islands (i.e., polders. This landscape is juxtaposed to the adjacent Sundarbans, a pristine mangrove forest, both well connected by a dense network of tidal channels that effectively convey water and sediment throughout the region. The extensive embanking in poldered areas, however, has greatly reduced the tidal prism (i.e., volume of water transported through local channels. We reveal that >600 km of these major waterways have infilled in recent decades, converting to land through enhanced sedimentation and the direct blocking of waterways by embankments and sluice gates. Nearly all of the observed closures (~98% have occurred along the embanked polder systems, with no comparable changes occurring in channels of the Sundarbans (<2% change. We attribute most of the channel infilling to the local reduction of tidal prism in poldered areas and the associated decline in current velocities. The infilled channels account for ~90 km2 of new land in the last 40–50 years, the rate of which, ~2 km2/yr, offsets the 4 km2/yr that is eroded at the coast, and is equivalent to ~20% of the new land produced naturally at the Ganges-Brahmaputra tidal rivermouth. Most of this new land, called ‘khas’ in Bengali, has been reclaimed for agriculture or aquaculture, contributing to the local economy. However, benefits are tempered by the loss of navigable waterways for commerce, transportation, and fishing, as well as the forced rerouting of tidal waters and sediments necessary to sustain this low-lying landscape against rising sea level. A more sustainable delta will require detailed knowledge of the consequences of these hydrodynamic changes to support more scientifically-grounded management of water, sediment, and tidal energy distribution.

46 CFR 129.220 - Basic safety.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Basic safety. 129.220 Section 129.220 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) OFFSHORE SUPPLY VESSELS ELECTRICAL INSTALLATIONS General Requirements § 129.220 Basic safety. (a) Electrical equipment and installations must be suitable...

Short and long term maintenance strategy for reactor vessel head penetrations

International Nuclear Information System (INIS)

Teissier, A.; Heuze, A.

1995-01-01

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

Reclamation of abandoned mined lands along th Upper Illinois Waterway using dredged material

Energy Technology Data Exchange (ETDEWEB)

Van Luik, A; Harrison, W

1982-01-01

Sediments were sampled and characterized from 28 actual or proposed maintenance-dredging locations in the Upper Illinois Waterway, that is, the Calumet-Sag Channel, the Des Plaines River downstream of its confluence with the Calumet-Sag Channel, and the Illinois River from the confluence of the Kankakee and Des Plaines rivers to Havana, Illinois. Sufficient data on chemical constituents and physical sediments were obtained to allow the classification of these sediments by currently applicable criteria of the Illinois Environmental Protection Agency for the identification of hazardous, persistent, and potentially hazardous wastes. By these criteria, the potential dredged materials studied were not hazardous, persistent, or potentially hazardous; they are a suitable topsoil/ reclamation medium. A study of problem abandoned surface-mined land sites (problem lands are defined as being acidic and/or sparsely vegetated) along the Illinois River showed that three sites were particularly well suited to the needs of the Corps of Engineers (COE) for a dredged material disposal/reclamation site. Thes sites were a pair of municipally owned sites in Morris, Illinois, and a small corporately owned site east of Ottawa, Illinois, and adjacent to the Illinois River. Other sites were also ranked as to suitability for COE involvement in their reclamation. Reclamation disposal was found to be an economically competitive alternative to near-source confined disposal for Upper Illinois Waterway dredged material.

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.)

Full vessel CFD analysis on thermal-hydraulic characteristics of CPR1000 PWR

International Nuclear Information System (INIS)

Chao Yanmeng; Yang Lixin; Zhang Mingqian

2014-01-01

To obtain flow distributions and thermal-hydraulic properties in a full vessel PWR under limited computation ability and time, a full vessel simulation model of CPR1000 was built based on two simplification methods. One simplified the inner geometry of the control rod guide tubes using equivalent flow area. Another substituted the core by a porous domain to maintain the pressure drop and temperature rise. After the computation, global and localized flow distributions, hydraulic loads of some main assemblies were obtained, as well as other thermal-hydraulic properties. The results indicate the flow distribution in the full vessel is asymmetrical. Therefore it is essential to use the full vessel model to simulate. The calculated thermal-hydraulic characteristics agree well with the operation statistics, providing the reference data for the reactor safety operation. (authors)

Push Off 2000 : new oilfield safety device catching on

Energy Technology Data Exchange (ETDEWEB)

Mowers, J.

2006-12-15

Fuel gas scrubbers use production gas to operate oil batteries, well separators, dehydrators, compressors, and pneumatic and pressure controls. Once an internal float gets stuck, production is stopped. The Push Off 2000 is a new safety device for fuel gas scrubbers which allows operators to easily dislodge a stuck internal float by activating the tool, which is mounted on top of the vessel. The device was developed after an operator suffered burns from a flash explosion that occurred after using a hammer to strike the fuel gas scrubber. The hammer method is the usual method of dealing with stuck internal floats, and can also jeopardize the integrity of the vessel, releasing gas or liquid hydrocarbons into the environment. The Push Off 2000 is expected to reduce costly facility downtime and increase revenue generating production time. Over 50 oil and gas companies have the Push Off 2000 device installed on fuel gas scrubber units at their facilities, and have recognized the safety and economic merits of the tool. The patented device has been approved by the Alberta Boilers Safety Association, the British Columbia Boiler and Elevator Safety Branch, and the Saskatchewan Boiler and Pressure Vessel Safety Unit. 1 fig.

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

Reliability-based dynamic positioning of floating vessels with riser and mooring system

DEFF Research Database (Denmark)

Fang, Shaoji; Leira, Bernt J.; Blanke, Mogens

2011-01-01

To maintain safety of a floating vessel with associated slender components such as risers and mooring line, the vessel is normally kept within a limited region. To specify a safe position in that region, this paper suggests a new position chasing algorithm with the consideration of both riser ang...... to their criticality. An optimal position set-point is produced by minimization of the value of the cost function. Numerical simulations show the effectiveness of the proposed algorithm....

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.

Assessment of Hydraulic Conditions Supporting the Recruitment of Asian Carp in the Illinois Waterway - A Case Study Using Known Spawning Events of 2015

Science.gov (United States)

Soong, D. T.; Garcia, T.; Duncker, J.; Zhu, Z.; Butler, S.; Diana, M.; Wahl, D.

2016-12-01

The upstream movement of Asian carp in the Illinois Waterway poses a potential threat to the Great Lakes. If established within the Great Lakes, Asian carp may disrupt the food web and harm the ecosystems of the Great Lakes. Understanding the Asian carp reproduction, including the timing and locations of adult spawning and the transport and dispersal of eggs and larvae, is essential information for managing the Asian carp population in the Illinois Waterway. The Fluvial Egg Drift Simulator (FluEgg) model, a Lagrangian particle tracking model, has been used to study the transport and dispersal of eggs and larvae. The FluEgg model inputs are water temperature and hydraulic properties. At present, field measured or modeled hydraulics from steady-state simulations have been used in FluEgg modeling and the applications have shown useful results for evaluating Asian carp reproduction in the Illinois Waterway. However, there is a need to use data based on more representative time-variable hydraulic conditions from spawning to the time larvae reach the Gas Bladder Inflation Stage (GBI). The GBI stage is critical because that is the stage when the young fish seek nursery habitat. In June 2015, Asian carp spawning was observed at two locations along the Illinois Waterway, one below Starved Rock Lock and Dam near Utica, and the one in the La Grange Pool near Havana, Illinois. This study analyzes how hydraulic modeling can improve the predictability of the FluEgg model. An unsteady HEC-RAS hydraulic model of the Illinois Waterway from Brandon Road Lock and Dam to Grafton, Illinois was used to reproduce the June 2015 flood event. Hydraulic data from HEC-RAS modeling, including predicted spatial and temporal discharge, water depth, and shear velocity; and measured water temperature data were used as input to the FluEgg model. FluEgg simulation results illustrate the downstream drifting of eggs and larvae until reaching the GBI stage. These simulation results can be analyzed

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

International Nuclear Information System (INIS)

Steffen, H.P.

1987-01-01

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

76 FR 11334 - Safety Zone; Soil Sampling; Chicago River, Chicago, IL

Science.gov (United States)

2011-03-02

...The Coast Guard is establishing a temporary safety zone on the North Branch of the Chicago River near Chicago, Illinois. This zone is intended to restrict vessels from a portion of the North Branch of the Chicago River due to soil sampling in this area. This temporary safety zone is necessary to protect the surrounding public and vessels from the hazards associated with the soil sampling efforts.

Longitudinal dispersion of radioactive substances in Federal waterways

Energy Technology Data Exchange (ETDEWEB)

Krause, W.J. [Bundesanstalt fuer Gewaesserkunde (BfG), Koblenz (Germany); Speer, W.; Luellwitz, T.; Cremer, M.; Tolksdorf, W.

2007-08-15

In the context of radioactivity monitoring in German Federal Waterways (BWStr) by the Federal Institute of Hydrology (BfG) according to the Precautionary Radiation Protection Act (StrVG), the prediction of the dispersion of radioactive substances in water is one of the key tasks. The aim is the forecasting of the longitudinal dispersion of concentrations of soluble hazardous substances in flowing water. These predictions are based on the so-called dispersion tests with tritium as a tracer that the BfG has performed since 1980. Characteristic parameters like discharge-dependent flow velocities, dispersion and elimination constants related to emission sources or selected river sections are determined. They will serve as basis for a mathematical model to forecast discharge-dependent flow velocities, expected impact times, concentration maxima, and the duration of critical increases in concentrations. In the following, the results obtained till now from three investigation campaigns on the River Weser and its source rivers Werra and Fulda are described. (orig.)

Evaluation of Agency Non-Code Layered Pressure Vessels (LPVs). Corrected Copy, Aug. 25, 2014

Science.gov (United States)

Prosser, William H.

2014-01-01

In coordination with the Office of Safety and Mission Assurance and the respective Center Pressure System Managers (PSMs), the NASA Engineering and Safety Center (NESC) was requested to formulate a consensus draft proposal for the development of additional testing and analysis methods to establish the technical validity, and any limitation thereof, for the continued safe operation of facility non-code layered pressure vessels. The PSMs from each NASA Center were asked to participate as part of the assessment team by providing, collecting, and reviewing data regarding current operations of these vessels. This report contains the outcome of the assessment and the findings, observations, and NESC recommendations to the Agency and individual NASA Centers.

46 CFR 176.910 - Passenger Ship Safety Certificate.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Passenger Ship Safety Certificate. 176.910 Section 176... 100 GROSS TONS) INSPECTION AND CERTIFICATION International Convention for Safety of Life at Sea, 1974, as Amended (SOLAS) § 176.910 Passenger Ship Safety Certificate. (a) A vessel, which carries more than...

Outlines of guidelines for the inspection and evaluation of reactor vessel internals

International Nuclear Information System (INIS)

Seki, Hiroaki; Kobayashi, Hiroyuki; Nakano, Morihito; Murai, Soutarou; Nomoto, Toshiharu

2014-01-01

'The guideline committee for the inspection and evaluation of Reactor Vessel Internals' of JANSI (Japan Nuclear Safety Institute) has been developing many guidelines based on principle which the conservative methodology, and covered both individual inspection method of reactor internals and application of repair methods for reactor internals. In this paper, some aspects of the JANSI-VIP-03 (Guidelines for the inspection and evaluation of Reactor Vessel Internals, revised Dec.2013) which is summary document of the committee activity, are introduced. (author)

Passive safety systems for decay heat removal of MRX

Energy Technology Data Exchange (ETDEWEB)

Ochiai, M; Iida, H; Hoshi, T [Japan Atomic Energy Research Inst., Ibaraki (Japan). Nuclear Ship System Lab.

1996-12-01

The MRX (marine Reactor X) is an advanced marine reactor, its design has been studied in Japan Atomic Energy Research Institute. It is characterized by four features, integral type PWR, in-vessel type control rod drive mechanisms, water-filled containment vessel and passive decay heat removal system. A water-filled containment vessel is of great advantage since it ensures compactness of a reactor plant by realizing compact radiation shielding. The containment vessel also yields passive safety of MRX in the event of a LOCA by passively maintaining core flooding without any emergency water injection. Natural circulation of water in the vessels (reactor and containment vessels) is one of key factors of passive decay heat removal systems of MRX, since decay heat is transferred from fuel rods to atmosphere by natural circulation of the primary water, water in the containment vessel and thermal medium in heat pipe system for the containment vessel water cooling in case of long terms cooling after a LOCA as well as after reactor scram. Thus, the ideal of water-filled containment vessel is considered to be very profitable and significant in safety and economical point of view. This idea is, however, not so familiar for a conventional nuclear system, so experimental and analytical efforts are carried out for evaluation of hydrothermal behaviours in the reactor pressure vessel and in the containment vessel in the event of a LOCA. The results show the effectiveness of the new design concept. Additional work will also be conducted to investigate the practical maintenance of instruments in the containment vessel. (author). 4 refs, 9 figs, 2 tabs.

Consequence evaluation of hypothetical reactor pressure vessel support failure

International Nuclear Information System (INIS)

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

1991-01-01

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

Safety margin evaluation of pre-stressed concrete nuclear containment vessel model with BARC code ULCA

International Nuclear Information System (INIS)

Basha, S.M.; Patnaik, R.; Ramanujam, S.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

2002-01-01

Full text: Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian pressurised heavy water reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results and for prediction of safety margins of Indian PHWRs. The present paper highlights the analysis results for prestressed concrete containment vessel (PCCV) tested at Sandia National Labs, USA in a round robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the most probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd = design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd

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)

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.

Summary of LWR safety research in the USA

International Nuclear Information System (INIS)

Murley, T.E.; Tong, L.S.; Bennett, G.L.

1977-01-01

The U.S. Nuclear Regulatory Commission's water reactor safety research program is described and the basic results are presented. The USNRC water reactor safety research program consists of five basic research areas: integrity of vessel and piping, thermal-hydraulic test, fuel rod behaviour, code development and verification, and reactor operational safety. Results from the vessel and piping integrity research have demonstrated the high safety margins in scaled vessels and the analytical procedures for calculating vessel behaviour under pressure. Non-destructive examination techniques are being improved. Work is also proceeding to define the material constituents to reduce the susceptibility of irradiation embrittlement and stress corrosion cracking. The thermal-hydraulic tests have covered the various phases of a hypothetical loss of coolant accident (LOCA) and activation of the emergency core cooling system (ECCS). These tests have led to the development of engineering correlations to describe the phenomena to further quantify the safety margins in commercial nuclear power plants. Specifically, this paper presents selected experimental data and analytical predictions from the initial tests in LOFT and SEMISCALE. Comparisons and evaluations are made between the data and analytical predictions. Significant results and conclusions are presented regarding the behaviour of emergency core cooling systems in a LOCA environment: the ability to predict LOCA-type experiments over a scaling range of thirty and the thermal-hydraulic behaviour of components such as pumps in an integral system LOCA environment. The fuel behaviour research has provided valuable information on decay heat, cladding oxidation, fuel rod behaviour and fuel metling. Both the decay heat and the cladding oxidation have been shown to be lower than assumed in the licensing evaluations. The fuel behaviour and thermo-hydraulic research is being integrated into computer codes to be used to provide additional

Probabilistic optimization of safety coefficients

International Nuclear Information System (INIS)

Marques, M.; Devictor, N.; Magistris, F. de

1999-01-01

This article describes a reliability-based method for the optimization of safety coefficients defined and used in design codes. The purpose of the optimization is to determine the partial safety coefficients which minimize an objective function for sets of components and loading situations covered by a design rule. This objective function is a sum of distances between the reliability of the components designed using the safety coefficients and a target reliability. The advantage of this method is shown on the examples of the reactor vessel, a vapour pipe and the safety injection circuit. (authors)

Proactive life extension of pressure vessels

Science.gov (United States)

Mager, Lloyd

1998-03-01

For a company to maintain its competitive edge in today's global market every opportunity to gain an advantage must be exploited. Many companies are strategically focusing on improved utilization of existing equipment as well as regulatory compliance. Abbott Laboratories is no exception. Pharmaceutical companies such as Abbott Laboratories realize that reliability and availability of their production equipment is critical to be successful and competitive. Abbott Laboratories, like many of our competitors, is working to improve safety, minimize downtime and maximize the productivity and efficiency of key production equipment such as the pressure vessels utilized in our processes. The correct strategy in obtaining these objectives is to perform meaningful inspection with prioritization based on hazard analysis and risk. The inspection data gathered in Abbott Laboratories pressure vessel program allows informed decisions leading to improved process control. The results of the program are reduced risks to the corporation and employees when operating pressure retaining equipment. Accurate and meaningful inspection methods become the cornerstone of a program allowing proper preventative maintenance actions to occur. Successful preventative/predictive maintenance programs must utilize meaningful nondestructive evaluation techniques and inspection methods. Nondestructive examination methods require accurate useful tools that allow rapid inspection for the entire pressure vessel. Results from the examination must allow the owner to prove compliance of all applicable regulatory laws and codes. At Abbott Laboratories the use of advanced NDE techniques, primarily B-scan ultrasonics, has provided us with the proper tools allowing us to obtain our objectives. Abbott Laboratories uses B-scan ultrasonics utilizing a pulse echo pitch catch technique to provide essential data on our pressure vessels. Equipment downtime is reduced because the nondestructive examination usually takes

77 FR 7025 - Safety Zones; America's Cup World Series, East Passage, Narragansett Bay, RI

Science.gov (United States)

2012-02-10

...-AA00 Safety Zones; America's Cup World Series, East Passage, Narragansett Bay, RI AGENCY: Coast Guard... the America's Cup World Series sailing vessel racing event. This safety zone is intended to safeguard...'s Cup-class races on the waters of the East Passage, Narragansett Bay, Rhode Island. Vessels will be...

Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

Energy Technology Data Exchange (ETDEWEB)

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

2011-10-15

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

Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

International Nuclear Information System (INIS)

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

2011-01-01

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

First evaluations of ex-vessel fuel-coolant interaction with MC3D

International Nuclear Information System (INIS)

Meignen, R.; Dupas, J.; Chaumont, B.

2003-01-01

In the frame of severe accident nuclear safety studies, we evaluate for French PWR's the potential of Steam Explosion in the reactor pit, consecutively to a vessel failure and to the mixing of the corium with the water that might be present. The evaluations are made with MC3D. This thermalhydraulic multiphasic code has firstly been qualified and its main parameters chosen so that a sufficient validation is obtained with regards to reactor situations. The safety study for ex-vessel situations is a step-by-step procedure that leads to a progressive process of hypotheses relaxations. We find that it is important to adequately model the corium ejection from the RPV. The rapid transition of the flow at the breach towards 2-phase dispersed flow leads to an important mixing of corium and water. The vessel pressurization is a very important parameter and strong pressure cases lead to a fine fragmentation and thus a high voiding. The small pressure cases are more dangerous for two reasons: the corium is dispersed in larger drops, and some important interactions (in the premixing sense) are reported

49 CFR 176.4 - Port security and safety regulations.

Science.gov (United States)

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Port security and safety regulations. 176.4... SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION HAZARDOUS MATERIALS REGULATIONS CARRIAGE BY VESSEL General § 176.4 Port security and safety regulations. (a) Each carrier, master, agent, and charterer of a...

Design safety improvements of Kozloduy NPP

International Nuclear Information System (INIS)

Hinovski, I.

1999-01-01

Design safety improvements of Kozloduy NPP, discussed in detail, are concerned with: primary circuit integrity; reactor pressure vessel integrity; primary coolant piping integrity; primary coolant overpressure protection; leak before break status; design basis accidents and transients; severe accident analysis; improvements of safety and support systems; containment/confinement leak tightness and strength; seismic safety improvements; WWER-1000 control rod insertion; upgrading and modernization of Units 5 and 6; Year 2000 problem

Research on the improvement of nuclear safety

International Nuclear Information System (INIS)

Yoo, Keon Joong; Kim, Dong Soo; Kim, Hui Dong; Park, Chang Kyu

1993-06-01

To improve the nuclear safety, this project is divided into three areas which are the development of safety analysis technology, the development of severe accident analysis technology and the development of integrated safety assessment technology. 1. The development of safety analysis technology. The present research aims at the development of necessary technologies for nuclear safety analysis in Korea. Establishment of the safety analysis technologies enables to reduce the expenditure both by eliminating excessive conservatisms incorporated in nuclear reactor design and by increasing safety margins in operation. It also contributes to improving plant safety through realistic analyses of the Emergency Operating Procedures (EOP). 2. The development of severe accident analysis technology. By the computer codes (MELCOR and CONTAIN), the in-vessel and the ex-vessel severe accident phenomena are simulated. 3. The development of integrated safety assessment technology. In the development of integrated safety assessment techniques, the included research areas are the improvement of PSA computer codes, the basic study on the methodology for human reliability analysis (HRA) and common cause failure (CCF). For the development of the level 2 PSA computer code, the basic research for the interface between level 1 and 2 PSA, the methodology for the treatment of containment event tree are performed. Also the new technologies such as artificial intelligence, object-oriented programming techniques are used for the improvement of computer code and the assessment techniques

Design features of the KSTAR in-vessel control coils

Energy Technology Data Exchange (ETDEWEB)

Kim, H.K. [National Fusion Research Institute (NFRI), 52 Yeoeun-dong, Yusung-ku, Daejeon, 305-333 (Korea, Republic of)], E-mail: hkkim@nfri.re.kr; Yang, H.L.; Kim, G.H.; Kim, Jin-Yong; Jhang, Hogun; Bak, J.S.; Lee, G.S. [National Fusion Research Institute (NFRI), 52 Yeoeun-dong, Yusung-ku, Daejeon, 305-333 (Korea, Republic of)

2009-06-15

In-vessel control coils (IVCCs) are to be used for the fast plasma position control, field error correction (FEC), and resistive wall mode (RWM) stabilization for the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The IVCC system comprises 16 segments to be unified into a single set to achieve following remarkable engineering advantages; (1) enhancement of the coil system reliability with no welding or brazing works inside the vacuum vessel, (2) simplification in fabrication and installation owing to coils being fabricated outside the vacuum vessel and installed after device assembly, and (3) easy repair and maintenance of the coil system. Each segment is designed in 8 turns coil of 32 mm x 15 mm rectangular oxygen free high conductive copper with a 7 mm diameter internal coolant hole. The conductors are enclosed in 2 mm thick Inconel 625 rectangular welded vacuum jacket with epoxy/glass insulation. Structural analyses were implemented to evaluate structural safety against electromagnetic loads acting on the IVCC for the various operation scenarios using finite element analysis. This paper describes the design features and structural analysis results of the KSTAR in-vessel control coils.

UK regulatory aspects of prestressed concrete pressure vessels for gas-cooled reactor nuclear power stations

International Nuclear Information System (INIS)

Watson, P.S.

1990-01-01

Safety assessment principles for nuclear power plants and for nuclear chemical plants demand application of best proven techniques, recognised standards, adequacy margins, inspection and maintenance of all the components including prestressed concrete pressure vessels. In service inspection of prestressed concrete pressure vessels includes: concrete surface examination; anchorage inspection; tendon load check; tendon material examination; foundation settlement and tilt; log-term deformation; vessel temperature excursions; coolant loss; top cap deflection. Hartlepool and Heysham 1 power plants prestress shortfall problem is discussed. Main recommendations can be summarised as follows: at all pressure vessel stations prestress systems should be calibrated in a manner which results in all load bearing components being loaded in a representative manner; at all pressure vessel stations load measurements during calibration should be verified by a redundant and diverse system

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

Analysis of cracked pressure vessel nozzles by finite elements

International Nuclear Information System (INIS)

Reynen, J.

1975-01-01

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

Ex-vessel coolability and energetics of steam explosions in nordic light water reactors

Energy Technology Data Exchange (ETDEWEB)

Park, H.S.; Dinh, T.N. [Royal Institute of Technology (Sweden)

2007-04-15

The report summarizes activities conducted at the Division of Nuclear Power Safety, Royal Institute of Technology-Sweden (KTH-NPS) within the ExCoolSe project during the year 2005, which is a transition year for the KTH-NPS program. The ExCoolSe project supported by NKS contributes to the severe accident research at KTH-NPS concurrently supported by APRI, HSK and EU SARNET. The main objective in ExCoolSe project is to scrutinize research on risk-significant safety issues related to severe accident management (SAM) strategy adopted for Nordic BWR plants, namely the Ex-vessel Coolability and Energetic Steam explosion. The work aims to pave way toward building a tangible research framework to tackle these long-standing safety issues. Chapter 1 describes the project objectives and work description. Chapter 2 provides a critical assessment of research results obtained from several past programs at KTH. This includes review of key data, insights and implications from POMECO (Porous Media Coolability) program, COMECO (Corium Melt Coolability) program, SIMECO (Study of In-Vessel Melt Coolability) program, and MISTEE (Micro-Interactions in Steam Explosion Experiments) program. Chapter 3 discusses the rationale of the new research program focusing on the SAM issue resolution. The program emphasizes identification and qualification of physics-based limiting mechanisms for both in-vessel phenomena (melt progression and debris coolability in the lower head, vessel failure), and ex-vessel phenomena. Chapter 4 introduces research results from the newly established DEFOR (Debris Formation) program and the ongoing MISTEE program. The focus of DEFOR is fulfill an apparent gap in the contemporary knowledge of severe accidents, namely mechanisms which govern the debris bed formation and bed characteristics. The later control the debris bed coolability. In the MISTEE program, methods for image synchronization and data processing were developed and tested, which enable processing of

Ex-vessel coolability and energetics of steam explosions in nordic light water reactors

International Nuclear Information System (INIS)

Park, H.S.; Dinh, T.N.

2007-04-01

The report summarizes activities conducted at the Division of Nuclear Power Safety, Royal Institute of Technology-Sweden (KTH-NPS) within the ExCoolSe project during the year 2005, which is a transition year for the KTH-NPS program. The ExCoolSe project supported by NKS contributes to the severe accident research at KTH-NPS concurrently supported by APRI, HSK and EU SARNET. The main objective in ExCoolSe project is to scrutinize research on risk-significant safety issues related to severe accident management (SAM) strategy adopted for Nordic BWR plants, namely the Ex-vessel Coolability and Energetic Steam explosion. The work aims to pave way toward building a tangible research framework to tackle these long-standing safety issues. Chapter 1 describes the project objectives and work description. Chapter 2 provides a critical assessment of research results obtained from several past programs at KTH. This includes review of key data, insights and implications from POMECO (Porous Media Coolability) program, COMECO (Corium Melt Coolability) program, SIMECO (Study of In-Vessel Melt Coolability) program, and MISTEE (Micro-Interactions in Steam Explosion Experiments) program. Chapter 3 discusses the rationale of the new research program focusing on the SAM issue resolution. The program emphasizes identification and qualification of physics-based limiting mechanisms for both in-vessel phenomena (melt progression and debris coolability in the lower head, vessel failure), and ex-vessel phenomena. Chapter 4 introduces research results from the newly established DEFOR (Debris Formation) program and the ongoing MISTEE program. The focus of DEFOR is fulfill an apparent gap in the contemporary knowledge of severe accidents, namely mechanisms which govern the debris bed formation and bed characteristics. The later control the debris bed coolability. In the MISTEE program, methods for image synchronization and data processing were developed and tested, which enable processing of

Geographic Information System (GIS) characterization of benthic and emergent areas in the Intracoastal Waterway, Sarasota County, Florida in 1987 (NODC Accession 0000607)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — This GIS layer graphically represents algae, seagrass, tidal marshes, mangroves, and oyster bed coverages found throughout the Intracoastal Waterway in Sarasota...

Investigation of impulsively loaded pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Brown, N.; Cornwell, R.; Hanner, D.; Leichter, H.; Mohr, P.

1963-10-15

Explosion containment vessels for containing from 2,000 to 3,000 five ton nuclear explosions are considered. Analysis methods appear adequate and lowest weights using the most advanced materials available in the next five years are projected.None of these materials can be fabricated today and all require extensive development. Present material technology limits the choice of materials and defines the weight. The addition of safety factors and fixtures (nozzles, etc.) will add to this weight considerably, and may well radically alter the vessel response. Improvements in the strength weight ratios of metals and glasses over those considered in this report do not appear reasonable at this time. Winding schemes to utilize the high strength of steel wires and somehow maintain a reasonable thickness appear to offer the most promise. A `ductile` beryllium would of course offer vast improvement, but no indications that this is being developed have appeared and all presently known beryllium is much too brittle.

Основні принципи проектування суден змішаного ріка-море плавання

OpenAIRE

Yehorov, Gennadiy V.; Yehorov, Aleksandr G.

2014-01-01

Existing design experience was based on river-sea going vessels (RSV) oriented for operation at European inland waterways ofRussia. Beginning from 2005 Marine Engineering Bureau (MEB) designed vessels for operation at mouse ports ofAzovSea; projects for Siberian rivers vessels were worked out. As a result the basic designing principles of dry-cargo vessels and tankers hulls of river-sea going type that have been built during 2001-2010 are considered. It is shown that all these vessels are des...

Radiation safety in sea transport of radioactive material in Japan

International Nuclear Information System (INIS)

Odano, N.; Yanagi, H.

2004-01-01

Radiation safety for sea transport of radioactive material in Japan has been discussed based on records of the exposed dose of sea transport workers and measured data of dose rate equivalents distribution inboard exclusive radioactive material shipping vessels. Recent surveyed records of the exposed doses of workers who engaged in sea transport operation indicate that exposed doses of transport workers are significantly low. Measured distribution of the exposed dose equivalents inboard those vessels indicates that dose rate equivalents inside those vessels are lower than levels regulated by the transport regulations of Japan. These facts clarify that radiation safety of inboard environment and handling of transport casks in sea transport of radioactive material in Japan are assured

Radiation safety in sea transport of radioactive material in Japan

Energy Technology Data Exchange (ETDEWEB)

Odano, N. [National Maritime Research Inst., Tokyo (Japan); Yanagi, H. [Nuclear Fuel Transport Co., Ltd., Tokyo (Japan)

2004-07-01

Radiation safety for sea transport of radioactive material in Japan has been discussed based on records of the exposed dose of sea transport workers and measured data of dose rate equivalents distribution inboard exclusive radioactive material shipping vessels. Recent surveyed records of the exposed doses of workers who engaged in sea transport operation indicate that exposed doses of transport workers are significantly low. Measured distribution of the exposed dose equivalents inboard those vessels indicates that dose rate equivalents inside those vessels are lower than levels regulated by the transport regulations of Japan. These facts clarify that radiation safety of inboard environment and handling of transport casks in sea transport of radioactive material in Japan are assured.

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

Prospects for nuclear safety research

Energy Technology Data Exchange (ETDEWEB)

Beckjord, E.S.

1995-04-01

This document is the text of a paper presented by Eric S. Beckjord (Director, Nuclear Regulatory Research/NRC) at the 22nd Water Reactor Safety Meeting in Bethesda, MD in October 1994. The following topics are briefly reviewed: (1) Reactor vessel research, (2) Probabilistic risk assessment, (3) Direct containment heating, (4) Advanced LWR research, (5) Nuclear energy prospects in the US, and (6) Future nuclear safety research. Subtopics within the last category include economics, waste disposal, and health and safety.

Question of right to stay of nuclear-powered vessels in foreign ports

International Nuclear Information System (INIS)

Szafranek, J.

1976-01-01

Question of right to stay of nuclear-powered vessels in foreign ports is considered in the light of London Convention on Safety of Life at Sea, Brussels Convention on Liability for Operation of Nuclear Ships and bilateral agreements. (Z.M.)

Advanced ultrasonic and eddy current examinations of the reactor vessel

International Nuclear Information System (INIS)

Cvitanovic, M.; Zado, V.

1996-01-01

In order to improve safety and reliability of nuclear power plant components, the existing examination methods are permanently developed as well as the new methods of examination are implemented. For the same reason, beside referent requirements, complementary NDE methods are utilized. Some examination methods techniques are not required to be used by referent safety codes and standards but they are frequently practiced as additional prevention to the component failure. This article presents the state of the art methods and techniques currently applied for examination of the reactor vessel base material, clad and weld materials. (author)

78 FR 10064 - Safety Zone; Vigor Industrial Roll-Out, West Duwamish Waterway, Seattle, WA

Science.gov (United States)

2013-02-13

... zone unless authorized by the Captain of the Port or a Designated Representative. DATES: This rule is... slow moving dry dock that will be maneuvering close to the shore, the Coast Guard is establishing a... cancelled sooner by the Captain of the Port or a Designated Representative. The safety zone will be enforced...

Quality assuring measures for pressure vessels - system approaches, certification, accreditation, surveillance

International Nuclear Information System (INIS)

Link, M.

1992-01-01

Quality assurance measures for pressure vessels in accordance with German codes and standards and with the participation of manufacturers, plant operators and third party inspection agencies represent a high standard in terms of engineering, safety and availability. Technical competence and the autonomous action of German industry in the field of quality assurance set internationally recognized safety standards. The continuous exchange of experience through the active involvement of manufacturers, plant operators and third party inspection agencies in work establishing codes and standards and in th updating of the state of the art give the German system a control loop and feedback function (Technical Committees on Pressure Vessels). Within the framework of European harmonization it is a German concern that technical competence and expertise are not lost in a formally legal, bureaucratic certification procedure. In the course of the European harmonization process, the dual German QA concept should maintain its position by utilizing the specialist knowledge and competence of experts, and permit appropriate adaptation. (orig.)

Twenty-fifth water reactor safety information meeting: Proceedings. Volume 1: Plenary sessions; Pressure vessel research; BWR strainer blockage and other generic safety issues; Environmentally assisted degradation of LWR components; Update on severe accident code improvements and applications

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following information: (1) plenary sessions; (2) pressure vessel research; (3) BWR strainer blockage and other generic safety issues; (4) environmentally assisted degradation of LWR components; and (5) update on severe accident code improvements and applications. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database.

Twenty-fifth water reactor safety information meeting: Proceedings. Volume 1: Plenary sessions; Pressure vessel research; BWR strainer blockage and other generic safety issues; Environmentally assisted degradation of LWR components; Update on severe accident code improvements and applications

International Nuclear Information System (INIS)

Monteleone, S.

1998-03-01

This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following information: (1) plenary sessions; (2) pressure vessel research; (3) BWR strainer blockage and other generic safety issues; (4) environmentally assisted degradation of LWR components; and (5) update on severe accident code improvements and applications. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database

Safety device of thermonuclear device

International Nuclear Information System (INIS)

Aoki, Isao; Ueda, Shuzo; Seki, Yasushi; Sakurai, Akiko; Kasahara, Fumio; Obara, Atsushi; Yamauchi, Michinori.

1997-01-01

The present invention provides a safety device against an event of intrusion of coolants in a vacuum vessel. Namely, a coolant supply system comprises cooling tubes for supplying coolants to main reactor structure components including a vacuum vessel. A detection means detects leakage of coolants in the vacuum vessel. A coolant supply control means controls the supply of coolants to the main reactor structural components based on the leakage detection signals of the detection means. A stagnated material discharging means discharges stagnated materials in the main reactor structural components caused by the leakage of coolants. The leakage of coolants (for example, water) in the vacuum vessel can thus be detected by the water detection device in the vacuum vessel. A control value of a coolant supply means is closed by the leakage detection signals. The supply of coolants to the main reactor structural components is restricted to suppress the leakage. The stagnated materials are discharged to a tank by way of a water draining valve. (I.S.)

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

CERN Document Server

American Society for Testing and Materials. Philadelphia

2007-01-01

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

Investigation of vessel exterior air cooling for a HLMC reactor

International Nuclear Information System (INIS)

Sienicki, J. J.; Spencer, B. W.

2000-01-01

The Secure Transportable Autonomous Reactor (STAR) concept under development at Argonne National Laboratory provides a small (300 MWt) reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100%+ natural circulation heat removal from the low power density/low pressure drop ultra-long lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the Reactor Exterior Cooling System (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the Reactor Vessel Auxiliary Cooling System (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink

Application of the ASME code in designing containment vessels for packages used to transport radioactive materials

International Nuclear Information System (INIS)

Raske, D.T.; Wang, Z.

1992-01-01

The primary concern governing the design of shipping packages containing radioactive materials is public safety during transport. When these shipments are within the regulatory jurisdiction of the US Department of Energy, the recommended design criterion for the primary containment vessel is either Section III or Section VIII, Division 1, of the ASME Boiler and Pressure Vessel Code, depending on the activity of the contents. The objective of this paper is to discuss the design of a prototypic containment vessel representative of a packaging for the transport of high-level radioactive material

Unstable fracture of nuclear pressure vessel

International Nuclear Information System (INIS)

Urata, Kazuyoshi

1978-01-01

Unstable fracture of nuclear pressure vessel shell for light water reactors up to 1,000 MWe class is discussed in accordance with ASME Code Sec. XI. The depth of surface crack required to protect against the unstable fracture is calculated on the basis of reactor operating conditions including loss of coolant accidents. Calculated surface crack depth a is equal to tαexp(2.19(a/l)) where l is crack length and t is weld thickness. α is crack depth required to protect against the unstable fracture in terms of the ratio of crack deth to weld thickness for surface crack have infinite length. Using this α, the safety factor included for allowable defect described in Sec. XI and the effects of thickness is discussed. It is derived that allowable defect described in Sec. XI include the safety factor of two on the crack depth for crack initiation at postulated accident and the safety factor of ten for crack depth calculated from point of view of crack arrest at normal conditions. (auth.)

The international state of affairs in marine safety

International Nuclear Information System (INIS)

Benkert, W.M.

1978-01-01

The three-fold objective of marine safety is examined with emphasis on international cooperation as a means of achievement. In this respect, the recent and present activities of the Intergovernmental Maritime Consultative organization are reviewed by looking at the accomplishments and goals of several subcommittees of the Maritime Safety Committee. The United States program for commercial vessel safety is briefly discussed along with a comment on the recent Tanker Safety initiatives

Patterns in benthic populations in the Milford Haven waterway following the 'Sea Empress' oil spill with special reference to amphipods.

Science.gov (United States)

Nikitik, Christopher C S; Robinson, Andrew W

2003-09-01

The macrobenthic fauna of the Milford Haven Waterway was studied in detail following the 'Sea Empress' oil spill in 1996. Contamination patterns indicated heaviest contamination of sediments by oil to have occurred in the lower reaches of the waterway, although water borne hydrocarbons are likely to have penetrated throughout the Haven. Generally, the communities showed little impact of contamination by oil, although some changes were evident at the population level. A decline in the amphipod fauna was observed throughout the Haven, with the genera Ampelisca and Harpinia and the family Isaeidae particularly affected. This was accompanied by increases in both the diversity and abundance of polychaete populations as opportunist species took advantage of the decline of the amphipod fauna. However, within five years of the spill the amphipod fauna has shown clear signs of recovery. The use of the polychaete/amphipod ratio as an indicator of oil pollution is discussed.

75 FR 12688 - Safety Zone; Gallants Channel, Beaufort, NC

Science.gov (United States)

2010-03-17

... and the use of heavy equipment and machinery in the waterway. Under 5 U.S.C. 553(d)(3), the Coast... debris and the use of heavy equipment and machinery in the waterway. Background and Purpose The State of... repairs present potential hazards to mariners and property from falling debris and the use of heavy...

LNG as a marine and inland waterway fuel. Contribution of AFG (French gas association) to the national policy framework for the deployment of alternative fuel infrastructures (AFNPF)

International Nuclear Information System (INIS)

2016-06-01

Here in the early 21. Century, the environmental footprint of marine and inland waterway transport is a major concern on a global scale. Under the authority of the International Maritime Organisation (IMO), regulations on atmospheric emissions are becoming increasingly tighter. The European Union (EU) is deeply committed to this endeavour and in particular has adopted Directive 2014/94/EU which aims to facilitate the deployment of an alternative fuels infrastructure. The Directive calls upon Member States to develop by 18 November 2016, national policy frameworks for the deployment of alternative fuel infrastructures (AFNPF) that must designate an appropriate number of LNG refuelling points in maritime and inland ports for maritime and inland waterway transport purposes. In light of this, the French Gas Association (AFG), in cooperation with all affected stakeholders and interested parties, has discussed the role of LNG as a marine and inland waterway fuel in the energy transition toward green growth, in order to contribute to the development of France's 'national policy framework' (AFNPF) under the Directive mentioned above. To assess the prospects of the LNG bunkering market in France, as well as the implementation of the corresponding infrastructures, the AFG adopted a scenario-based approach from the user's perspective. Considering that we are in an early stage, the study focuses on the types of ships that are most likely to use LNG, and it examines their traffic in France's major metropolitan ports, along with forecasted activity based on the strategic plans for those ports. Two scenarios - one being rather conservative (baseline) and the other being more optimistic - were examined over three timelines: 2020, 2025 and 2030. The scenarios lead to a gradual implementation of LNG fuel in French ports, on the three coastlines and along the five inland waterway corridors. They show LNG fuel needs in all the ports of the central TEN

Public safety around dams : Ontario Power Generation's approach

Energy Technology Data Exchange (ETDEWEB)

Bennett, T; Rowat, L [Ontario Power Generation Inc., Toronto, ON (Canada)

2009-04-01

Ontario Power Generation (OPG) has developed a Waterways Public Safety Program that includes elements such as integrating public safety considerations into normal business practices and decisions; applying conservative decision making principles regarding operations where there are issues of public safety; and seeking partnership opportunities that enhance public safety awareness. The key steps to the public safety risk assessment process include identifying each type of known public interaction at a facility; identifying the hazards associated with those interactions; assigning a rating of likelihood and consequence for each separate public interaction at a facility; and assigning a risk rating, with a risk matrix for each public interaction. OPG now requires that a new public safety risk assessment be completed every 3 years. The risk assessment is a guide to implementing control measures to lower the risk of injury at dam facilities. OPG has adopted the model that every water conveyance structure will have an established hazardous area, typically adjacent to the structure. These areas are delineated with red danger signs and other control measures installed as needed. Yellow signs are used to delineate warning areas where there is a reasonable likelihood of minor injury or where the public may become stranded. As a minimum, OPG uses signage, sluicegate audible alarms, a stepped approach to sluicegate openings and public education. Safety booms and buoys as well as fencing and barricades may be used as additional control measures along with security patrols and video surveillance to target specific public interactions. 6 figs.

From automatic control of vessels designed to operate in offshore conditions to automatic control of a large barge convoy in a river/channel: lessons learned; Do controle automatico de embarcacoes projetadas para condicoes operacionais de offshore para o controle automatico de um grande comboio de barcacas em um rio/canal: licoes aprendidas

Energy Technology Data Exchange (ETDEWEB)

Pontes, Anselmo Carvalho [Symmetry Ltda., Rio de Janeiro, RJ (Brazil); Santos, Edson Mesquita do [Marinha do Brasil, Rio de Janeiro, RJ (Brazil)

2008-07-01

The waterway transportation sector, despite its problems and bottlenecks, is continually seeking alternatives to improve its operations. With goals of increased efficiency, safety and cost reduction, the sector is modernizing and acquiring new technology. The Dynamic Position control system for a large barge convoy is one of the new technology highlights. This innovation may make development in the Brazilian waterway transportation sector feasible by permitting navigational safety and greater flexibility in transport and distribution not only for mined products but for petroleum and its derivatives. This paper presents the development process of The Dynamic Position control system for a large barge convoy in the Parana-Paraguay River, for Rio Tinto Mining Company. The hydrodynamics effects in a river are quite different than those of the traditional offshore environment. The degree of difficulty involved in solving this problem is highly dependent on a ship's characteristics, which can be analyzed in function of their hydrodynamic coefficients. Model tests in towing tanks were made to adjust the hydrodynamic coefficients and characteristics of the thrusters. The choice of the 'backstepping' control law and observer (state estimator) presented in this paper is a result of these particular conditions. The authors point out the advantage of using this technique, since all the properties of the coefficients of the hull, propeller(s) and thruster(s) have already been tested and tuned and are able to make a reliable set-up that optimizes navigation in restricted waterways. (author)

A powerful methodology for reactor vessel pressurized thermal shock analysis

International Nuclear Information System (INIS)

Boucau, J.; Mager, T.

1994-01-01

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

78 FR 36790 - Commercial Fishing Safety Advisory Committee; Vacancies

Science.gov (United States)

2013-06-19

... position); (c) a representative of education or training professionals related to fishing vessel safety or...) manufacturers of equipment for vessels to which Chapter 45 of Title 46, U.S.C. applies; (3) education or... Homeland Security (DHS) does not discriminate in employment on the basis of race, color, religion, sex...

Application of Kalman filter to short-term tide level prediction by Yen, P.-H., Jan, C.-D., Lee, Y.-P. and Lee, H.-F. (J. Waterway Port Coast. Ocean Eng. 122(5); 1996; pp 226-231: Discussion

Digital Repository Service at National Institute of Oceanography (India)

Mandal, S.; Chandramohan, P.

stream_size 2 stream_content_type text/plain stream_name J_Waterway_Port_Coast_Eng_124_213.pdf.txt stream_source_info J_Waterway_Port_Coast_Eng_124_213.pdf.txt Content-Encoding ISO-8859-1 Content-Type text/plain; charset=ISO-8859-1 ...

Regulatory Assessment Technologies for Aging of Reactor Vessel Internals

Energy Technology Data Exchange (ETDEWEB)

Jhung, Myung Jo; Park, Jeong Soon; Ko, Hanok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

In order to develop the audit calculation system, it is required to develop crack evaluation, seismic analysis and thermal-hydraulic analysis techniques for RVIs so that integrity of RVIs under the aging environment can be evaluated and be assured. In addition, regulatory requirements including safety review and inspection guides should be developed in order to assure the quality and uniformity of safety reviews and inspections regarding aging assessment and management of RVIs. Since Reactor Vessel Internals (RVIs) are installed within the reactor pressure vessel and surround the fuel assemblies, some of them are exposed to the environment such as high neutron irradiation, high temperature and reactor coolant flow. Those environmental factors can cause damage to RVIs including cracks, loss of material, fatigue, loss of fracture toughness and change of dimension as the operation time of nuclear power plants (NPPs) increases. For long-term operation more than 40 years, aging management of RVIs is important. The final objectives of this study are to establish the audit calculation system for RVIs and to develop regulatory requirements for aging assessment and management of RVIs considering their operating conditions, materials, and possible aging mechanisms.

Fusion reactor safety studies, FY 1977

International Nuclear Information System (INIS)

Darby, J.B. Jr.

1978-04-01

This report reviews the technical progress in the fusion reactor safety studies performed during FY 1977 in the Fusion Power Program at the Argonne National Laboratory. The subjects reported on include safety considerations of the vacuum vessel and first-wall design for the ANL/EPR, the thermal responses of a tokamak reactor first wall, the vacuum wall electrical resistive requirements in relationship to magnet safety, and a major effort is reported on considerations and experiments on air detritiation

Materials surveillance program for C-E NSSS reactor vessels

International Nuclear Information System (INIS)

Koziol, J.J.

1977-01-01

Irradiation surveillance programs for light water NSSS reactor vessels provide the means by which the utility can assess the extent of neutron-induced changes in the reactor vessel materials. These programs are conducted to verify, by direct measurement, the conservatism in the predicted radiation-induced changes and hence the operational parameters (i.e., heat-up, cooldown, and pressurization rates). In addition, such programs provide assurance that the scheduled adjustments in the operational parameters are made with ample margin for safe operation of the plant. During the past 3 years, several documents have been promulgated establishing the criteria for determining both the initial properties of the reactor vessel materials as well as measurement of changes in these initial properties as a result of irradiation. These documents, ASTM E-185-73, ''Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels,'' and Appendix H to 10 CFR 50, ''Reactor Vessel Material Surveillance Program Requirements,'' are complementary to each other. They are the result of a change in the basic philosophy regarding the design and analysis of reactor vessels. In effect, the empirical ''transition temperature approach,'' which was used for design, was replaced by the ''analytical fracture mechanics approach.'' The implementation of this technique was described in Welding Research Council Bulletin 1975 and Appendix G to ASME Code Section III. Further definition of requirements appears in Appendix G to 10 CFR 50 published in July 1973. It is the intent of this paper to describe (1) a typical materials surveillance program for the reactor vessel of a Combustion Engineering NSSS, and (2) how the results of such programs, as well as experimental programs provide feed-back for improvement of materials to enhance their radiation resistance and thereby further improve the safety and reliability of future plants. (author)

Revisiting the analysis of passive plasma shutdown during an ex-vessel loss of coolant accident in ITER blanket

International Nuclear Information System (INIS)

Rivas, J.C.; Dies, J.; Fajarnés, X.

2015-01-01

Highlights: • We have repeated the safety analysis for the hypothesis of passive plasma shutdown for beryllium evaporation during an ex-vessel LOCA of ITER first wall, with AINA code. • We have performed a sensitivity analysis over some key parameters that represents uncertainties in physics and engineering, to identify cliff edge effects. • The obtained results for the 500 MW inductive scenario, with an ex-vessel LOCA affecting a third of first wall surface are similar to those of previous studies and point to the possibility of a passive plasma shutdown during this safety case, before a serious damage is inflicted to the ITER wall. • The sensitivity analysis revealed a new scenario potentially damaging for the first wall if we increase fusion power and time delay for impurity transport, and decrease fraction of affected first wall area and initial beryllium fraction in plasma. • After studying the 700 MW inductive scenario, with an ex-vessel LOCA affecting 10% of first wall surface, with 0.5% of Be in plasma and a time delay twice the energy confinement time, it was found that affected area of first wall would melt before a passive plasma shutdown occurs. - Abstract: In this contribution, the analysis of passive safety during an ex-vessel loss of coolant accident (LOCA) in the first wall/shield blanket of ITER has been studied with AINA safety code. In the past, this case has been studied using robust safety arguments, based on simple 0D models for plasma balance equations and 1D models for wall heat transfer. The conclusion was that, after first wall heating up due to the loss of all coolant, the beryllium evaporation in the wall surface would induce a growing impurity flux into core plasma that finally would end in a passive shut down of the discharge. The analysis of plasma-wall transients in this work is based in results from AINA code simulations. AINA (Analyses of IN vessel Accidents) code is a safety code developed at Fusion Energy Engineering

76 FR 1362 - Safety Zone; Ice Conditions for the Baltimore Captain of Port Zone

Science.gov (United States)

2011-01-10

... hazards include vessels becoming beset or dragged off course, sinking or grounding, and creating hazards... safety zone's intended objectives of protecting persons and vessels from becoming beset or dragged off... there is little vessel traffic associated with recreational boating and commercial fishing during the...

Electromagnetic wave survey on voids behind waterway channel lining; Suiro kaikyo sokuheki haimen kudo no denjiha tansa

Energy Technology Data Exchange (ETDEWEB)

Koitabashi, H [Tokyo Electric Power Co. Inc., Tokyo (Japan); Inagaki, M

1996-10-01

Voids behind lining were surveyed by applying electromagnetic wave reflection method to the waterway channel of a hydraulic power plant. Since waterway channel lining is ranged from oblique to vertical direction, voids are hardly formed. However, formation of voids or cavities behind lining is supposed such as voids between ground and lining due to change with time or consolidation settlement, and voids due to soil loss. Electromagnetic radar reflection suggesting continuous void was observed behind terrace concrete lining. As the result of core boring, thin continuous void of 2-5cm thick and more than 100m long was found. This was possibly formed by consolidation settlement for a long time. In some sites, continuous void signal was observed at the upper part of side walls although this signal was smaller than that at the upper part of a terrace. This continuous cavity of 10-20cm thick and 20m long was different from voids, and unevenly distributed at the upper part of an open channel along flowing surface with large flow rate. In addition, it is necessary to clarify the relation to cracks. 2 refs., 4 figs.

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

International Nuclear Information System (INIS)

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

2012-11-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-11-15

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

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.

Optimal support vessel and access system selection for offshore wind farms

NARCIS (Netherlands)

Asgarpour, M.; Serraris, J.W.; Ridder, E. de; Broek, J. van den; Bos, J.E.

2016-01-01

Operation and maintenance (O&M) costs contribute to a significant part of Cost of Energy produced by offshore wind. In order to reduce the O&M costs and to guarantee safety and wellbeing of the maintenance technicians, for each individual wind farm an optimal set of support vessels and access

Composite Overwrap Pressure Vessels: Mechanics and Stress Rupture Lifting Philosophy

Science.gov (United States)

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

2009-01-01

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

Composite Overwrap Pressure Vessels: Mechanics and Stress Rupture Lifing Philosophy

Science.gov (United States)

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

2007-01-01

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

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

International Nuclear Information System (INIS)

1998-10-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. A separate abstract was prepared for the introduction and for each of the eleven chapters, which are: 1. Reactor Pressure Vessel Design, 2. Reactor Pressure Materials, 3. WWER Pressure Vessels, 4. Determination of Mechanical Properties, 5. Neutron Exposure, 6. Methodology of Irradiation Experiments, 7. Effect of Irradiation on Mechanical Properties, 8. Mechanisms of Irradiation Embrittlement, 9. Modelling of Irradiation Damage, 10. Annealing of Irradiation Damage, 11. Safety Assessment using Surveillance Programmes and Data Bases

A classification system for pressure vessel shell failures

International Nuclear Information System (INIS)

Harrop, L.P.

1989-01-01

A system for classifying failures of the shells of pressure vessels is presented. The classification system is based on the way a failure physically manifests itself and not on imputed economic or safety significance. It is believed the described way of classifying the failures is useful for transferring information from one situation to another. In assigning names to types of failure, the intention has been to adopt explicit definitions rather than supposed colloquial usage. (author)

Re-examining reactor vessel embrittlement at Chooz A

International Nuclear Information System (INIS)

Guilleret, J.-C.

1988-01-01

The Chooz A PWR experienced an extended shutdown in 1987/88 following indications that the reactor vessel was embrittling more rapidly than expected. Discrepancies between the expected rate and estimates of the actual rate were not easily explained. The huge body of work done since then to establish safety margins and support restart of the plant should provide a model for the owners of other older PWRs grappling with the embrittlement issue. (author)

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)

Life extension of the St. Lucie unit 1 reactor vessel

International Nuclear Information System (INIS)

Rowan, G.A.; Sun, J.B.; Mott, S.L.

1991-01-01

In late 1989, Florida Power and Light Company (FP and L) established the policy that St. Lucie unit 1 should not be prevented from achieving a 60-yr operating life by reactor vessel embrittlement. A 60-yr operating life means that the plant would be allowed to operate until the year 2036, which is 20 years beyond the current license expiration date of 2016. Since modifications to the reactor vessel and its components are projected to be expensive, the desire of FP and L management was to achieve this lifetime extension through the use of fuel management and proven technology. The following limitations were placed on any acceptable method for achieving this lifetime extension capability: low fuel cycle cost; low impact on safety parameters; very little or no operations impact; and use of normal reactor materials. A task team was formed along with the Advanced Nuclear Fuels Company (ANF) to develop a vessel-life extension program

Possible research program on a large scale nuclear pressure vessel

International Nuclear Information System (INIS)

1983-01-01

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

Stent-assisted coil embolization of aneurysms with small parent vessels: safety and efficacy analysis.

Science.gov (United States)

Kühn, Anna Luisa; Hou, Samuel Y; Puri, Ajit S; Silva, Christine F; Gounis, Matthew J; Wakhloo, Ajay K

2016-06-01

Stent-assisted coil embolization (SACE) is a viable therapeutic approach for wide-neck intracranial aneurysms. However, it can be technically challenging in small cerebral vessels (≤2 mm). To present our experience with stents approved for SACE in aneurysms with small parent arteries. All patients who underwent stent-assisted aneurysm treatment with either a Neuroform or an Enterprise stent device at our institution between June 2006 and October 2012 were identified. Additionally, we evaluated each patient's vascular risk factors, aneurysm characteristics (ruptured vs non-ruptured, incidental finding, recanalized) and follow-up angiography data. A total of 41 patients with 44 aneurysms met our criteria, including 31 women and 10 men. Most of the aneurysms were located in the anterior circulation (75%). Stent placement in vessels 1.2-2 mm in diameter was successful in 93.2%. Thromboembolic complications occurred in 6 cases and vessel straightening was seen in 1 case only. Initial nearly complete to complete aneurysm obliteration was achieved in 88.6%. Six-month follow-up angiography showed coil compaction in three cases, one asymptomatic in-stent stenosis and stent occlusion. Twelve to 20-months' follow-up showed stable coil compaction in two patients compared with previous follow-up, and aneurysm recanalization in two patients. Twenty-four to 36-months' follow-up showed further coil compaction in one of these patients and aneurysm recanalization in a previous case of stable coil compaction on mid-term follow-up. Our results suggest that SACE of aneurysms with small parent vessels is feasible in selected cases and shows good long-term patency rates of parent arteries. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/

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

Implementation of Council Directive 96/29 EURATOM. Notice to owners, operators and managers of ships, inland waterways craft and fishing vessels, masters, officers and ratings of merchant ships, and skippers and crew of fishing vessels and persons in charge of inland waterway craft

International Nuclear Information System (INIS)

2001-01-01

This Marine Guidance Note draws attention to the obligations under several statutory regulations to ensure compliance with Council Directive 96/29 Euratom of 13 May 1996 laying down basic safety standards for the protection and health of workers and the general public against the dangers arising from ionising radiation. (author)

Structural integrity evaluation of PWR nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Cruz, Julio R.B.; Mattar Neto, Miguel

1999-01-01

The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

75 FR 8431 - Carbon Dioxide Fire Suppression Systems on Commercial Vessels

Science.gov (United States)

2010-02-24

... and nonsubstantive style, format, or wording changes that we are proposing solely to improve the... 1. The authority citation for Part 25 continues to read as follows: Authority: 33 U.S.C. 1903(b); 46... Commanding Officer, U.S. Coast Guard Marine Safety Center. PART 27--TOWING VESSELS 3. The authority citation...

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

International Nuclear Information System (INIS)

Guelicher, L.; Schilling, F.E.

1977-01-01

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

Review of in-service thermal annealing of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Server, W.L.

1984-01-01

Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper-shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. A test reactor pressure vessel has been wet annealed at less than 343 0 C (650 0 F), and annealing of the Belgian BR-3 reactor vessel has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place is feasible, but solvable engineering problems do exist. The materials with highest radiation sensitivity in the older reactor vessels are submerged-arc weld metals with high copper and nickel concentrations. The limited Charpy V-notch and fracture toughness data available for five such welds were reviewed. The review suggested that significant recovery results from annealing at 454 0 C (850 0 F) for one week. Two of the main concerns with a localized heat treatment at 454 0 C (850 0 F) are the degree of distortion that may occur after the annealing cycle and the extent of residual stresses. A thermal and structural analysis of a reactor vessel for distortions and residual stresses found no problems with the reactor vessel itself but did indicate a rotation at the nozzle region of the vessel that would plastically deform the attached primary piping. Further analytical studies are needed. An American Society for Testing and Materials (ASTM) task group is upgrading and revising the ASTM Recommended Guide for In-Service Annealing of WaterCooled Nuclear Reactor Vessels (E 509-74) with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (for example, the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

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...

Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

International Nuclear Information System (INIS)

Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

1996-01-01

Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs