Review of advanced reactor transient analysis capabilities and applications for Savannah River Plant reactors

International Nuclear Information System (INIS)

Buckner, M.R.; Hostetler, D.E.; Anderson, M.M.; Dodds, H.L.

1977-01-01

GRASS is a three-dimensional, coupled neutronic and engineering code for analysis of the radioisotope production reactors at the Savannah River Plant. The capabilities of GRASS are reviewed with emphasis on recent additions to model accident conditions involving the transport of molten fuel material and to accurately characterize neutronic and engineering feedback. The general application of GRASS to the Savannah River reactors is discussed, and results are presented for the analyses of severla reactor transient calculations

Experimental effect of flow depth on ratio discharge in lateral intakes in river bend

International Nuclear Information System (INIS)

Masjedi, A; Foroushani, E P

2012-01-01

Open-channel dividing flow is characterized by the inflow and outflow discharges, the upstream and downstream water depths, and the recirculation flow in the branch channel. In general, diversion flow can be categorized as natural and artificial flow. Natural flow diversion usually occurs as braiding or cut-off in bend rivers, while artificial flow is man-made to divert flow by lateral intake channels for water supply. This study presents the results of a laboratory research into effect intake flow depth on ratio discharge in lateral intakes in 180 degree bend. Investigation on lateral intake and determination of intake flow depth is among the most important issues in lateral intake on ratio discharge with model intake flow depth were measured in a laboratory flume under clear-water. Experiments were conducted for various intake flow depths and with different discharges. It was found that by increasing the flow depth at 180 degree flume bend, ratio discharge increases.

Advantages of customer/supplier involvement in the upgrade of River Bend's IST program

International Nuclear Information System (INIS)

Womack, R.L.; Addison, J.A.

1996-01-01

At River Bend Station, IST testing had problems. Operations could not perform the test with the required repeatability; engineering could not reliably trend test data to detect degradation; licensing was heavily burdened with regulatory concerns; and maintenance could not do preventative maintenance because of poor prediction of system health status. Using Energy's Total Quality principles, it was determined that the causes were: lack of ownership, inadequate test equipment usage, lack of adequate procedures, and lack of program maintenance. After identifying the customers and suppliers of the IST program data, Energy management put together an upgrade team to address these concerns. These customers and suppliers made up the IST upgrade team. The team's mission was to supply River Bend with a reliable, functional, industry correct and user friendly IST program. The IST program in place went through a verification process that identified and corrected over 400 individual program discrepancies. Over 200 components were identified for improved testing methods. An IST basis document was developed. The operations department was trained on ASME Section XI testing. All IST tests have been simplified and shortened, due to heavy involvement by operations in the procedure development process. This significantly reduced testing time, resulting in lower cost, less dose and greater system availability

A (desintegração da África pós-colonial em A Bend in the River de V. S. Naipaul = The (disintegration of post-colonial Africa in A Bend in the River by V. S. Naipaul

Directory of Open Access Journals (Sweden)

Mariana Bolfarine

2010-01-01

Full Text Available O presente trabalho visa refletir sobre as consequências do imperialismo europeu em uma cidade africana fictícia, inspirada no Zaire, representada por V. S. Naipaul na obra A Bend in The River (1979. O artigo investiga a falta de integração entre as diferentesesferas sociais que se formaram no leste da África, a partir da imigração de indianos e asiáticos como indentured workers, trabalhadores contratados. Serão analisados trechos do romance que possuem como foco as consequências dos relacionamentos inter-raciaisvividos entre diferentes personagens. A conclusão é que A Bend in the River demonstra que a política de dividir para governar, posta em prática pelos europeus, influenciou a construção de uma sociedade fragmentada, desigual, hierárquica e determinista. O pessimismo deNaipaul corrobora a ideologia racista do colonialismo, que prega a pressuposição pelo negro africano da superioridade do branco europeu, reafirmando o seu direito de oprimir e dominar os povos colonizados. O campo teórico é constituído com base nos estudos de Avtar Brah, Edward Said, Michael Gorra.This work aims at reflecting upon the consequences of European imperialism in a fictitious African city, supposedly situated in Zaire, represented by V. S. Naipaul in A Bend in the River (1979. The article investigates the lack of integration between different social spheres, which were formed in East Africa, after the immigration of Indians and Asians as indentured workers. Different excerpts of the novel that focus on the consequences of the inter-racial relationships between different characters in the narrative will be analyzed. The conclusion is that A Bend in the River reveals that the divide and rule policy, put into practice by the Europeancolonizer, influenced the construction of a society that is fragmented, unequal, hierarchical and deterministic. Naipaul’s pessimism supports the racist ideology of colonialism, which preachesthe assumption by the

Conformance to Regulatory Guide 1.97, River Bend Station, Unit No. 1 (Docket No. 50-458)

International Nuclear Information System (INIS)

Udy, A.C.

1985-08-01

This EG and G, Inc., report reviews the submittals for Regulatory Guide 1.97, Revision 3, for the River Bend Station, Unit No. 1. Any exception to Regulatory Guide 1.97 is evaluated and those areas where sufficient basis for acceptability is not provided are identified. 8 refs

76 FR 81992 - PPL Bell Bend, LLC; Combined License Application for Bell Bend Nuclear Power Plant; Exemption

Science.gov (United States)

2011-12-29

... License Application for Bell Bend Nuclear Power Plant; Exemption 1.0 Background PPL Bell Bend, LLC... for Nuclear Power Plants.'' This reactor is to be identified as Bell Bend Nuclear Power Plant (BBNPP... based upon the U.S. EPR reference COL (RCOL) application for UniStar's Calvert Cliffs Nuclear Power...

Population trends, bend use relative to available habitat and within-river-bend habitat use of eight indicator species of Missouri and Lower Kansas River benthic fishes: 15 years after baseline assessment

Science.gov (United States)

Wildhaber, Mark L.; Yang, Wen-Hsi; Arab, Ali

2016-01-01

A baseline assessment of the Missouri River fish community and species-specific habitat use patterns conducted from 1996 to 1998 provided the first comprehensive analysis of Missouri River benthic fish population trends and habitat use in the Missouri and Lower Yellowstone rivers, exclusive of reservoirs, and provided the foundation for the present Pallid Sturgeon Population Assessment Program (PSPAP). Data used in such studies are frequently zero inflated. To address this issue, the zero-inflated Poisson (ZIP) model was applied. This follow-up study is based on PSPAP data collected up to 15 years later along with new understanding of how habitat characteristics among and within bends affect habitat use of fish species targeted by PSPAP, including pallid sturgeon. This work demonstrated that a large-scale, large-river, PSPAP-type monitoring program can be an effective tool for assessing population trends and habitat usage of large-river fish species. Using multiple gears, PSPAP was effective in monitoring shovelnose and pallid sturgeons, sicklefin, shoal and sturgeon chubs, sand shiner, blue sucker and sauger. For all species, the relationship between environmental variables and relative abundance differed, somewhat, among river segments suggesting the importance of the overall conditions of Upper and Middle Missouri River and Lower Missouri and Kansas rivers on the habitat usage patterns exhibited. Shoal and sicklefin chubs exhibited many similar habitat usage patterns; blue sucker and shovelnose sturgeon also shared similar responses. For pallid sturgeon, the primary focus of PSPAP, relative abundance tended to increase in Upper and Middle Missouri River paralleling stocking efforts, whereas no evidence of an increasing relative abundance was found in the Lower Missouri River despite stocking.

Westinghouse independent safety review of Savannah River production reactors

International Nuclear Information System (INIS)

Leggett, W.D.; McShane, W.J.; Liparulo, N.J.; McAdoo, J.D.; Strawbridge, L.E.; Call, D.W.

1989-01-01

Westinghouse Electric Corporation has performed a safety assessment of the Savannah River production reactors (K, L, and P) as requested by the US Department of Energy. This assessment was performed between November 1, 1988, and April 1, 1989, under the transition contract for the Westinghouse Savannah River Company's preparations to succeed E.I. du Pont de Nemours ampersand Company as the US Department of Energy contractor for the Savannah River Project. The reviewers were drawn from several Westinghouse nuclear energy organizations, embody a combination of commercial and government reactor experience, and have backgrounds covering the range of technologies relevant to assessing nuclear safety. The report presents the rationale from which the overall judgment was drawn and the basis for the committee's opinion on the phased restart strategy proposed by E.I. du Pont de Nemours ampersand Company, Westinghouse, and the US Department of Energy-Savannah River. The committee concluded that it could recommend restart of one reactor at partial power upon completion of a list of recommended upgrades both to systems and their supporting analyses and after demonstration that the organization had assimilated the massive changes it will have undergone. 37 refs., 1 fig., 3 tabs

Westinghouse independent safety review of Savannah River production reactors

Energy Technology Data Exchange (ETDEWEB)

Leggett, W.D.; McShane, W.J. (Westinghouse Hanford Co., Richland, WA (USA)); Liparulo, N.J.; McAdoo, J.D.; Strawbridge, L.E. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear and Advanced Technology Div.); Toto, G. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear Services Div.); Fauske, H.K. (Fauske and Associates, Inc., Burr Ridge, IL (USA)); Call, D.W. (Westinghouse Savannah R

1989-04-01

Westinghouse Electric Corporation has performed a safety assessment of the Savannah River production reactors (K,L, and P) as requested by the US Department of Energy. This assessment was performed between November 1, 1988, and April 1, 1989, under the transition contract for the Westinghouse Savannah River Company's preparations to succeed E.I. du Pont de Nemours Company as the US Department of Energy contractor for the Savannah River Project. The reviewers were drawn from several Westinghouse nuclear energy organizations, embody a combination of commercial and government reactor experience, and have backgrounds covering the range of technologies relevant to assessing nuclear safety. The report presents the rationale from which the overall judgment was drawn and the basis for the committee's opinion on the phased restart strategy proposed by E.I. du Pont de Nemours Company, Westinghouse, and the US Department of Energy-Savannah River. The committee concluded that it could recommend restart of one reactor at partial power upon completion of a list of recommended upgrades both to systems and their supporting analyses and after demonstration that the organization had assimilated the massive changes it will have undergone.

Final environmental statement related to the operation of River Bend Station (Docket No. 50-458)

International Nuclear Information System (INIS)

1985-01-01

This Final Environmental Statement contains the second assessment of the environmental impact associated with the operation of River Bend Station, pursuant to the National Environment Policy Act of 1969 (NEPA) and Title 10 of the Code of Federal Regulations, Part 51, as amended, of the Nuclear Regulatory Commission regulations. This statement examines the environment, environmental consequences and mitigating actions, and environmental and economic benefits and costs

Draft environmental statement related to the operation of River Bend Station (Docket No. 50-458)

International Nuclear Information System (INIS)

1984-07-01

This draft environmental statement contains the second assessment of the environmental impact associated with the operation of River Bend Station, pursuant to the National Environmental Policy Act of 1969 (NEPA) and Title 10 of the Code of Federal Regulations, Part 51, as amended, of the Nuclear Regulatory Commission regulations. This statement examines the environment, environmental consequences and mitigating actions, and environmental and economic benefits and costs

Savannah River Site reactor hardware design modification study

International Nuclear Information System (INIS)

Fisher, J.E.

1990-01-01

A study was undertaken to assess the merits of proposed design modifications to the Savannah River Site (SRS) reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. System recovery potential was evaluated for break locations at the pump suction, the pump discharge, and the plenum inlet. The code version used was RELAP5/MOD2.5 version 3d3, a preliminary version of RELAP5/MOD3. The model was a three-dimensional representation of the K-Reactor water plenum and moderator tank. It included explicit representations of all six loops, which were based on the configuration of L-Reactor. A combination of features is recommended to ensure liquid inventory recovery for all break locations. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit from the clamshell enclosing that section of piping. 7 refs., 10 figs., 2 tabs

Savannah River Site K-Reactor Probabilistic Safety Assessment

International Nuclear Information System (INIS)

Brandyberry, M.D.; Bailey, R.T.; Baker, W.H.; Kearnaghan, D.P.; O'Kula, K.R.; Wittman, R.S.; Woody, N.D.; Amos, C.N.; Weingardt, J.J.

1992-12-01

This report gives the results of a Savannah River Site (SRS) K-Reactor Probabilistic Safety Assessment (PSA). Measures of adverse consequences to health and safety resulting from representations of severe accidents in SRS reactors are presented. In addition, the report gives a summary of the methods employed to represent these accidents and to assess the resultant consequences. The report is issued to provide useful information to the U. S. Department of Energy (DOE) on the risk of operation of SRS reactors, for insights into severe accident phenomena that contribute to this risk, and in support of improved bases for other DOE programs in Heavy Water Reactor safety

Chemical aspects of gadolinium nitrate as soluble nuclear poison in Savannah River Plant reactors

International Nuclear Information System (INIS)

Baumann, E.W.

1978-01-01

The aqueous solution chemistry of gadolinium nitrate was studied to identify conditions that interfere with successful cleanup of gadolinium in Savannah River Plant reactor systems. Injecting a gadolinium nitrate solution into the D 2 O coolant-moderator constitutes a supplementary mode of reactor shutdown. The resulting approximately 0.001M gadolinium nitrate solution is then deionized by recirculation through mixed-bed ion exchange resins before reactor operation is resumed

EXPERIMENTAL EVALUATION OF THE FULLY LOADED ELK RIVER REACTOR

Energy Technology Data Exchange (ETDEWEB)

Fisher, J. R.; Diaz, A.

1963-06-15

The loading and testing program of the Elk River Reactor confirmed the predicted values. The measured cold, clean excess reactivity agrees to 2% and the control rod worths to 1% of the calculated values. The reactivity for various core loadings and rod positions is tabulated. The effects of spiked elements on the reactivity and radial peak-toaverage power ratio were studied. (D.L.C.)

Rivers running deep : complex flow and morphology in the Mahakam River, Indonesia

NARCIS (Netherlands)

Vermeulen, B.

2014-01-01

Rivers in tropical regions often challenge our geomorphological understanding of fluvial systems. Hairpin bends, natural scours, bifurcate meander bends, tie channels and embayments in the river bank are a few examples of features ubiquitous in tropical rivers. Existing observation techniques

Restart of K-Reactor, Savannah River Site: Safety evaluation report

Energy Technology Data Exchange (ETDEWEB)

1991-04-01

This Safety Evaluation Report (SER) focuses on those issues required to support the restart of the K-Reactor at the Savannah River Plant. This SER provides the safety criteria for restart and documents the results of the staff reviews of the DOE and operating contractor activities to meet these criteria. To develop the restart criteria for the issues discussed in this SER, the Savannah River Restart Office and Savannah River Special Projects Office staffs relied, when possible, on commercial industry codes and standards and on NRC requirements and guidelines for the commercial nuclear industry. However, because of the age and uniqueness of the Savannah River reactors, criteria for the commercial plants were not always applicable. In these cases, alternate criteria were developed. The restart criteria applicable to each of the issues are identified in the safety evaluations for each issue. The restart criteria identified in this report are intended to apply only to restart of the Savannah River reactors. Following the development of the acceptance criteria, the DOE staff and their support contractors evaluated the results of the DOE and operating contractor (WSRC) activities to meet these criteria. The results of those evaluations are documented in this report. Deviations or failures to meet the requirements are either justified in the report or carried as open or confirmatory items to be completed and evaluated in supplements to this report before restart. 62 refs., 1 fig.

Restart of K-Reactor, Savannah River Site: Safety evaluation report

International Nuclear Information System (INIS)

1991-04-01

This Safety Evaluation Report (SER) focuses on those issues required to support the restart of the K-Reactor at the Savannah River Plant. This SER provides the safety criteria for restart and documents the results of the staff reviews of the DOE and operating contractor activities to meet these criteria. To develop the restart criteria for the issues discussed in this SER, the Savannah River Restart Office and Savannah River Special Projects Office staffs relied, when possible, on commercial industry codes and standards and on NRC requirements and guidelines for the commercial nuclear industry. However, because of the age and uniqueness of the Savannah River reactors, criteria for the commercial plants were not always applicable. In these cases, alternate criteria were developed. The restart criteria applicable to each of the issues are identified in the safety evaluations for each issue. The restart criteria identified in this report are intended to apply only to restart of the Savannah River reactors. Following the development of the acceptance criteria, the DOE staff and their support contractors evaluated the results of the DOE and operating contractor (WSRC) activities to meet these criteria. The results of those evaluations are documented in this report. Deviations or failures to meet the requirements are either justified in the report or carried as open or confirmatory items to be completed and evaluated in supplements to this report before restart. 62 refs., 1 fig

Reactor safety research and development in Chalk River Laboratories

Energy Technology Data Exchange (ETDEWEB)

Nitheanandan, T. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

Atomic Energy of Canada Limited's Chalk River Laboratories provides three different services to stakeholders and customers. The first service provided by the laboratory is the implementation of Research and Development (R&D) programs to provide the underlying technological basis of safe nuclear power reactor designs. A significant portion of the Canadian R&D capability in reactor safety resides at Atomic Energy of Canada Limited's Chalk River Laboratories, and this capability was instrumental in providing the science and technology required to aid in the safety design of CANDU power reactors. The second role of the laboratory has been in supporting nuclear facility licensees to ensure the continued safe operation of nuclear facilities, and to develop safety cases to justify continued operation. The licensing of plant life extension is a key industry objective, requiring extensive research on degradation mechanisms, such that safety cases are based on the original safety design data and valid and realistic assumptions regarding the effect of ageing and management of plant life. Recently, Chalk River Laboratories has been engaged in a third role in research to provide the technical basis and improved understanding for decision making by regulatory bodies. The state-of-the-art test facilities in Chalk River Laboratories have been contributing to the R&D needs of all three roles, not only in Canada but also in the international community, thorough Canada's participation in cooperative programs lead by International Atomic Energy Agency and the OECD's Nuclear Energy Agency. (author)

Techniques for processing remote field eddy current signals from bend regions of steam generator tubes of prototype fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Thirunavukkarasu, S. [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India); Rao, B.P.C., E-mail: bpcrao@igcar.gov.in [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India); Jayakumar, T.; Raj, Baldev [Non Destructive Evaluation Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)

2011-04-15

Steam generator (SG) is one of the most critical components of sodium cooled fast breeder reactor. Remote field eddy current (RFEC) technique has been chosen for in-service inspection (ISI) of these ferromagnetic SG tubes made of modified 9Cr-1Mo steel (Grade 91). Expansion bends are provided in the SGs to accommodate differential thermal expansion. During ISI using RFEC technique, in expansion bend regions, exciter-receiver coil misalignment, bending stresses, probe wobble and magnetic permeability variations produce disturbing noise hindering detection of defects. Fourier filtering, cross-correlation and wavelet transform techniques have been studied for noise reduction as well as enhancement of RFEC signals of defects in bend regions, having machined grooves and localized defects. Performance of these three techniques has been compared using signal-to-noise ratio (SNR). Fourier filtering technique has shown better performance for noise reduction while cross-correlation technique has resulted in significant enhancement of signals. Wavelet transform technique has shown the combined capability of noise reduction and signal enhancement and resulted in unambiguous detection of 10% of wall loss grooves and localized defects in the bend regions with SNR better than 7 dB.

Savannah River Site production reactor technical specifications. K Production Reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-02-01

These technical specifications are explicit restrictions on the operation of the Savannah River Site K Production Reactor. They are designed to preserve the validity of the plant safety analysis by ensuring that the plant is operated within the required conditions bounded by the analysis, and with the operable equipment that is assumed to mitigate the consequences of an accident. Technical specifications preserve the primary success path relied upon to detect and respond to accidents. This report describes requirements on thermal-hydraulic limits; limiting conditions for operation and surveillance for the reactor, power distribution control, instrumentation, process water system, emergency cooling and emergency shutdown systems, confinement systems, plant systems, electrical systems, components handling, and special test exceptions; design features; and administrative controls.

Reactor loops at Chalk River

International Nuclear Information System (INIS)

Sochaski, R.O.

1962-07-01

This report describes broadly the nine in-reactor loops, and their components, located in and around the NRX and NRU reactors at Chalk River. First an introduction and general description is given of the loops and their function, supplemented with a table outlining some loop specifications and nine simplified flow sheets, one for each individual loop. The report then proceeds to classify each loop into two categories, the 'main loop circuit' and the 'auxiliary circuit', and descriptions are given of each circuit's components in turn. These components, in part, are comprised of the main loop pumps, the test section, loop heaters, loop coolers, delayed-neutron monitors, surge tank, Dowtherm coolers, loop piping. Here again photographs, drawings and tables are included to provide a clearer understanding of the descriptive literature and to include, in tables, some specifications of the more important components in each loop. (author)

Clinch River Breeder Reactor Plant: a building block in nuclear technology

International Nuclear Information System (INIS)

McCormack, M.

1979-01-01

Interest in breeder reactors dates from the Manhatten Project to the present effort to build the Clinch River Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. Seven breeder-type reactors which were built during this time are described and their technological progress assessed. The Clinch River Breeder Reactor Project (CRBRP) has been designed to demonstrate that it can be licensed, can operate on a large power grid, and can provide industry with important experience. As the next logical step in LMFBR development, the project has suffered repeated cancellation efforts with only minor modifications to its schedule. Controversies have developed over the timing of a large-scale demonstration plant, the risks of proliferation, economics, and other problems. Among the innovative developments adopted for the CRBRP is a higher thermal efficiency potential, the type of development which Senator McCormack feels justifies continuing the project. He argues that the nuclear power program can and should be revitalized by continuing the CRBRP

Clinch River Breeder Reactor Plant: a building block in nuclear technology

Energy Technology Data Exchange (ETDEWEB)

McCormack, M.

1979-01-01

Interest in breeder reactors dates from the Manhatten Project to the present effort to build the Clinch River Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. Seven breeder-type reactors which were built during this time are described and their technological progress assessed. The Clinch River Breeder Reactor Project (CRBRP) has been designed to demonstrate that it can be licensed, can operate on a large power grid, and can provide industry with important experience. As the next logical step in LMFBR development, the project has suffered repeated cancellation efforts with only minor modifications to its schedule. Controversies have developed over the timing of a large-scale demonstration plant, the risks of proliferation, economics, and other problems. Among the innovative developments adopted for the CRBRP is a higher thermal efficiency potential, the type of development which Senator McCormack feels justifies continuing the project. He argues that the nuclear power program can and should be revitalized by continuing the CRBRP.

L-Reactor operation, Savannah River Plant: environmental assessment

International Nuclear Information System (INIS)

1982-08-01

The purpose of this document is to assess the significance of the effects on the human environment of the proposed resumption of L-reactor operation at the Savannah River Plant, scheduled for October 1983. The discussion is presented under the following section headings: need for resumption of L-Reactor operations and purpose of this environmental assessment; proposed action and alternative; affected environment (including, site location and description, land use, historic and archeological resources, socioeconomic and community characteristics, geology and seismology, hydrology, meteorology and climatology, ecology, and radiation environment); environmental consequences; summary of projected L-Reactor releases and impacts; and Federal and State permits and approval. The three appendices are entitled: radiation dose calculation methods and assumptions; floodplain/wetlands assessment - L-Reactor operations; and, conversion table. A list of references is included at the end of each chapter

Computer program for modelling the history of the in-service bending of fast power reactor fuel assemblies

International Nuclear Information System (INIS)

Dienstbier, J.

1979-04-01

The studies into stresses and deformations in the core are mainly focused on the fuel rod and the fuel assembly can. In high neutron doses austenitic steel swells and this is associated with a considerable increase in the volume of material. The SANDRA computer program is used for solving the problems of can deformations and stress during long-term reactor operation. The block for the mechanical interaction of cans is the key part of the program. The program input data include temperature distribution, fast neutron flux distribution and coolant overpressure inside the cans. Reactor operation is modelled using operating modes A, B, C which may arbitrarily be combined. Mode A computes bending deformations and the deformations of the can cross-section due to temperature dilatation in the change in temperature fields in the reactor; mode B computes deformations due to swelling and creep in long-term operation; mode C computes thermal deformations in reactor shut-down. A flowsheet is shown of program SANDRA as are examples of computed deformations. (M.S.)

Field Investigation of Flow Structure and Channel Morphology at Confluent-Meander Bends

Science.gov (United States)

Riley, J. D.; Rhoads, B. L.

2007-12-01

The movement of water and sediment through drainage networks is inevitably influenced by the convergence of streams and rivers at channel confluences. These focal components of fluvial systems produce a complex hydrodynamic environment, where rapid changes in flow structure and sediment transport occur to accommodate the merging of separate channel flows. The inherent geometric and hydraulic change at confluences also initiates the development of distinct geomorphic features, reflected in the bedform and shape of the channel. An underlying assumption of previous experimental and theoretical models of confluence dynamics has been that converging streams have straight channels with angular configurations. This generalized conceptualization was necessary to establish confluence planform as symmetrical or asymmetrical and to describe subsequent flow structure and geomorphic features at confluences. However, natural channels, particularly those of meandering rivers, curve and bend. This property and observation of channel curvature at natural junctions have led to the hypothesis that natural stream and river confluences tend to occur on the concave outer bank of meander bends. The resulting confluence planform, referred to as a confluent-meander bend, was observed over a century ago but has received little scientific attention. This paper examines preliminary data on three-dimensional flow structure and channel morphology at two natural confluent-meander bends of varying size and with differing tributary entrance locations. The large river confluence of the Vermilion River and Wabash River in west central Indiana and the comparatively small junction of the Little Wabash River and Big Muddy Creek in southeastern Illinois are the location of study sites for field investigation. Measurements of time-averaged three-dimensional velocity components were obtained at these confluences with an acoustic Doppler current profiler for flow events with differing momentum ratios. Bed

Quality assurance in technology development for The Clinch River Breeder Reactor Plant Project

International Nuclear Information System (INIS)

Anderson, J.W.

1980-01-01

The Clinch River Breeder Reactor Plant Project is the nation's first large-scale demonstration of the Liquid Metal Fast Breeder Reactor (LMFBR) concept. The Project has established an overall program of plans and actions to assure that the plant will perform as required. The program has been established and is being implemented in accordance with Department of Energy Standard RDT F 2-2. It is being applied to all parts of the plant, including the development of technology supporting its design and licensing activity. A discussion of the program as it is applied to development is presented

Measuring device for bending of beryllium reflector

International Nuclear Information System (INIS)

Nishida, Seiri; Sakamoto, Naoki.

1994-01-01

The device of the present invention can measure bending of a beryllium reflector formed in a reactor core of a nuclear reactor by a relatively easy operation. Namely, a sensor portion comprises a long-support that can be inserted to a fuel element-insertion hole disposed in the reactor and a plurality of distance sensors disposed in a longitudinal direction of the support. A supersonic wave sensor which is advantageous in the heat resistance, the size and the accuracy and can conduct measurement in water relatively easily is used as the distance sensors. However, other sensors, instead of the sensor described above, may also be used. The plurality of distance sensors detect the bending amount of the beryllium reflector in the longitudinal direction by such an easy operation of inserting such a sensor portion to the fuel element-insertion hole upon exchange of fuel elements. (I.S.)

Flow visualization study of two-phase flow in a single bend outlet feeder pipe of a CANDU reactor

International Nuclear Information System (INIS)

Savalaxs, S.-A.; Lister, D.H.; Steward, F.R.

2005-01-01

In CANDU reactors, the feeder piping that is used to direct the high-temperature water coolant between the fuel channels and the steam generators is made of carbon steel. Since 1996, several CANDU stations have reported excessive corrosion of their outlet feeders. The first metre is particularity vulnerable because the piping there consists of single or double bends, which have relatively thin walls produced by the bending process. Early studies related the attack to the hydrodynamics of the coolant and verified that it was a type of flow-accelerated corrosion. In order to understand the hydrodynamics of the coolant in the outlet feeders by flow visualization, a full-scale transparent test section simulating the geometry and orientation of an outlet feeder bend with its upstream components was fabricated. The feeder consisted of a 54 mm diameter acrylic pipe with a 73 degree bend. This was connected to the upstream component with an acrylic simulation of a Grayloc flanged fitting. A test loop supplied room temperature water to the test section at flow rates up to 0.019 m3/s. Air could be injected into the water to give a mean volume fraction of up to 0.56. In this preliminary investigation, the size and velocity of air bubbles at different flow conditions and their distribution within the pipe bend were studied. Particular attention was paid to the flow pattern at the inside of the bend, where a CFD (computational fluid dynamics) code - Fluent 6.1-had failed to predict a liquid film in an earlier study. A high-speed digital video camera was used to determine the relation between bubble size and velocity. Such a relation should help to explain the discrepancy in the CFD modelling and provide the basis for accurate predictions of phase distribution in complex geometries at high flow rates. (authors)

Spontaneous bending of 2D molecular bottle-brush

NARCIS (Netherlands)

Subbotin, A; Jong, J; ten Brinke, G

Using a scaling approach we consider a 2D comb copolymer brush under bending deformations. We show that the rectilinear brush is locally stable and can be characterized by a persistence length lambda increasing with the molecular weight of grafting side chains as lambda similar to M-3. A bending

Monte Carlo verification of control-rod worth for the Savannah River K reactor

International Nuclear Information System (INIS)

Mosteller, R.D.

1992-01-01

The Savannah River K Reactor is a heavy-water reactor that relies on control-rod movement to control its reactivity and power distribution during normal operations. It is necessary, therefore, to have an accurate estimate of the reactivity worth of its control rods in order to predict the behavior of the reactor. Westinghouse Savannah River Company (WSRC) uses the GLASS lattice-physics code to calculate few-group cross sections for fuel and control-rod assemblies in the K reactor. This paper compares the control-rod worth calculated by GLASS to that calculated by the MCNP Monte Carlo program. The GLASS calculations utilize its standard 37-group cross-section library, while the MCNP calculations employ continuous-energy isotopic cross-section libraries derived from ENDF/B-V. The MCNP calculations therefore combine the most rigorous analytical model and the most accurate cross sections currently available for thermal-reactor analysis. Consequently, the MCNP results comprise a computational benchmark against which the accuracy of the GLASS code can be evaluated

The creep bending of short radius pipe bends

International Nuclear Information System (INIS)

Spence, John

1975-01-01

In existing and proposed liquid metal fast breeder reactor design the pipework has considerable importance. Parts of the LMFBR include thin walled short radius bends which are expected to operate in the creep regime. In linear elasticity it is known that the assumption of long radius bends is not too severe as far as the flexibility characteristics are concerned although some modifications are necessary for accurate determination of the stresses. No data exists for nonlinear creep. Current work is aimed at elucidating the effect of the various assumptions common to linear elastic theory in so far as they affect the creep characteristics of bends on systems. Herein an energy based analysis using a simple n power constitutive law for stationary creep is employed to derive basic design data for flexibilities and stresses which will be necessary before complete systems can be assessed for creep. The analysis shows on comparison with the long radius work that the assumption of R>r is not much more restrictive in creep than for linear elasticity. Flexibilities for short radius bends appear to be well approximated by the long radius values. Thus the attractive reference stress information already derived may be used directly to find deformations without a complete knowledge of the constitutive relationship. However, stresses are somewhat different. Fortunately the maximum deviation occurs at relatively low levels of stress, the peak stresses being in fair agreement. When n=1 the present results reduce essentially to those obtained from existing linear elastic theory

External events analysis for the Savannah River Site K reactor

International Nuclear Information System (INIS)

Brandyberry, M.D.; Wingo, H.E.

1990-01-01

The probabilistic external events analysis performed for the Savannah River Site K-reactor PRA considered many different events which are generally perceived to be ''external'' to the reactor and its systems, such as fires, floods, seismic events, and transportation accidents (as well as many others). Events which have been shown to be significant contributors to risk include seismic events, tornados, a crane failure scenario, fires and dam failures. The total contribution to the core melt frequency from external initiators has been found to be 2.2 x 10 -4 per year, from which seismic events are the major contributor (1.2 x 10 -4 per year). Fire initiated events contribute 1.4 x 10 -7 per year, tornados 5.8 x 10 -7 per year, dam failures 1.5 x 10 -6 per year and the crane failure scenario less than 10 -4 per year to the core melt frequency. 8 refs., 3 figs., 5 tabs

Clinch River Breeder Reactor Plant Project: construction schedule

International Nuclear Information System (INIS)

Purcell, W.J.; Martin, E.M.; Shivley, J.M.

1982-01-01

The construction schedule for the Clinch River Breeder Reactor Plant and its evolution are described. The initial schedule basis, changes necessitated by the evaluation of the overall plant design, and constructability improvements that have been effected to assure adherence to the schedule are presented. The schedule structure and hierarchy are discussed, as are tools used to define, develop, and evaluate the schedule

Discussion about design basis flood of site of research reactors by river

International Nuclear Information System (INIS)

Rong Feng; Zhao Jianjun; Du Qiaomin; Zhang Lingyan

2006-01-01

This paper presents the well-defined standard in relation to design the basis flood of the sites of research reactors by river. It is based on the concept of some relational standards, analysis of hydrological calculation technology and methods, and analysis of accident dangerous degrees of research reactor, as well as in combination with the engineering practices. The flood preventing standard for research reactors with higher power should be the same with that of the nuclear power plants. (authors)

Clinch river breeder reactor plant steam generator water quality

Energy Technology Data Exchange (ETDEWEB)

Van Hoesen, D; Lowe, P A

1975-07-01

The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: (1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; (2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and (3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present.

Clinch river breeder reactor plant steam generator water quality

International Nuclear Information System (INIS)

Van Hoesen, D.; Lowe, P.A.

1975-01-01

The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: 1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; 2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and 3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present

Flow Structure and Channel Morphology at a Confluent-Meander Bend

Science.gov (United States)

Riley, J. D.; Rhoads, B. L.

2009-12-01

Flow structure and channel morphology in meander bends have been well documented. Channel curvature subjects flow through a bend to centrifugal acceleration, inducing a counterbalancing pressure-gradient force that initiates secondary circulation. Transverse variations in boundary shear stress and bedload transport parallel cross-stream movement of high velocity flow and determine spatial patterns of erosion along the outer bank and deposition along the inner bank. Laboratory experiments and numerical modeling of confluent-meander bends, a junction planform that develops when a tributary joins a meandering river along the outer bank of a bend, suggest that flow and channel morphology in such bends deviate from typical patterns. The purpose of this study is to examine three-dimensional (3-D) flow structure and channel morphology at a natural confluent-meander bend. Field data were collected in southeastern Illinois where Big Muddy Creek joins the Little Wabash River near a local maximum of curvature along an elongated meander loop. Measurements of 3-D velocity components were obtained with an acoustic Doppler current profiler (ADCP) for two flow events with differing momentum ratios. Channel bathymetry was also resolved from the four-beam depths of the ADCP. Analysis of velocity data reveals a distinct shear layer flanked by dual helical cells within the bend immediately downstream of the confluence. Flow from the tributary confines flow from the main channel along the inner part of the channel cross section, displacing the thalweg inward, limiting the downstream extent of the point bar, protecting the outer bank from erosion and enabling bar-building along this bank. Overall, this pattern of flow and channel morphology is quite different from typical patterns in meander bends, but is consistent with a conceptual model derived from laboratory experiments and numerical modeling.

Bending force constant of gamma-ray irradiated NaNO2

International Nuclear Information System (INIS)

Kwun, S.I.; Allavena, M.

1976-01-01

The origin of the new peak appearing near the ν 2 i.r. absorption band of the NO 2 - group in γ-ray irradiated NaNO 2 ferroelectric crystal is explained by using a model which assumes that some of the Na + ions are displaced from their original sites after irradiation, perturbing the vibrational motion of NO 2 - . In this framework, the bending force constant of the perturbed NO 2 - group is calculated using a modified version of the CNDO/2 method, which can take into account the environmental effects on the local crystal site considered. The values of the bending force constant of virginal and irradiated NaNO 2 obtained are 1.19 md/A and 1.27 md/A respectively. The vibrational bending mode of the perturbed NO 2 - groups seems responsible for the additional i.r. absorption band observed experimentally at 835 cm -1 . (author)

Use of digital computers in the protection system for Savannah River reactors

International Nuclear Information System (INIS)

Gimmy, K.L.

1977-06-01

Each production reactor at the Savannah River Plant has recently been provided with a protective system using dual digital computers. The dual ''safety computers'' monitor coolant temperature and flow in each of the 600 fuel assemblies in the reactor. The system provides alarms and automatic reactor shutdown (SCRAM) if these variables exceed predetermined setpoints. The system provides the primary protection for unwanted local or general power increase or assembly coolant flow reduction. Standard process control computers are used and all scanning, data output, and protective action are controlled by software prepared by Du Pont

Functional safeguards for computers for protection systems for Savannah River reactors

International Nuclear Information System (INIS)

Kritz, W.R.

1977-06-01

Reactors at the Savannah River Plant have recently been equipped with a ''safety computer'' system. This system utilizes dual digital computers in a primary protection system that monitors individual fuel assembly coolant flow and temperature. The design basis for the (SRP safety) computer systems allowed for eventual failure of any input sensor or any computer component. These systems are routinely used by reactor operators with a minimum of training in computer technology. The hardware configuration and software design therefore contain safeguards so that both hardware and human failures do not cause significant loss of reactor protection. The performance of the system to date is described

Characterization of 2D-C/C composite for application of very high temperature reactor

International Nuclear Information System (INIS)

Shibata, Taiju; Sumita, Junya; Kunimoto, Eiji; Sawa, Kazuhiro; Makita, Taiyo; Takagi, Takashi; Kim, W.J.; Jung, C.H.; Park, J.Y.

2010-01-01

For in-core components of VHTR (Very High Temperature Reactor), carbon fiber reinforced carbon matrix composite (C/C composite) is one of the major candidate materials. In this study, fracture behaviors of two dimensional (2D-) C/C composites were examined by SENB specimens with four-point bending test. The surface of specimens was observed by a CCD camera during the bending test, and observed by a stereomicroscope before and after the bending test. The following results were obtained through mode-I fracture test. (1) Three types of the composites were evaluated by tentatively using the stress intensity factor equation for metallic materials. The equivalent stress intensity factor of 2D-C/C composite is in the range of 5.9 - 10.0MPa m 1/2 . It was expected that the fracture mechanism for the composite materials could be assessed by this test method. (2) The crack opening displacement-load behavior of C/C composite might depend not only on the propagation of crack but also on delaminating between layers. (author)

On the Simulation of Floods in a Narrow Bending Valley: The Malpasset Dam Break Case Study

Directory of Open Access Journals (Sweden)

Chiara Biscarini

2016-11-01

Full Text Available In this paper, we investigate the performance of three-dimensional (3D hydraulic modeling when dealing with river sinuosity and meander bends. In river bends, the flow is dominated by a secondary current, which has a key role on the flow redistribution. The secondary flow induces transverse components of the bed shear stress and increases the velocity in outward direction, thus generating local erosion and riverbed modifications. When in river bends, the 3D processes prevail, and a 3D computational fluid dynamics (CFD model is required to correctly predict the flow structure. An accurate description of the different hydrodynamic processes in mildly and sharply curved bends find a relevant application in meanders migration modeling. The mechanisms that drive the velocity redistribution in meandering channels depend on the river’s roughness, the flow depth (H, the radius curvature (R, the width (B and the bathymetric variations. Here, the hydro-geomorphic characterization of sharp and mild meanders is performed by means of the ratios R/B, B/H, and R/H, and of the sinuosity index. As a case study, we selected the Malpasset dam break on the Reyran River Valley (FR, as it is perfectly suited for investigating performances and issues of a 3D model in simulating the inundation dynamics in a river channel with a varying curvature radius.

Reliability modeling of Clinch River breeder reactor electrical shutdown systems

International Nuclear Information System (INIS)

Schatz, R.A.; Duetsch, K.L.

1974-01-01

The initial simulation of the probabilistic properties of the Clinch River Breeder Reactor Plant (CRBRP) electrical shutdown systems is described. A model of the reliability (and availability) of the systems is presented utilizing Success State and continuous-time, discrete state Markov modeling techniques as significant elements of an overall reliability assessment process capable of demonstrating the achievement of program goals. This model is examined for its sensitivity to safe/unsafe failure rates, sybsystem redundant configurations, test and repair intervals, monitoring by reactor operators; and the control exercised over system reliability by design modifications and the selection of system operating characteristics. (U.S.)

GRIMH3: A new reactor calculation code at Savannah River Site

International Nuclear Information System (INIS)

Le, T.T.; Pevey, R.E.

1993-01-01

The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex. The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex

Finite Element Analysis for Bending Process of U-Bending Specimens

Energy Technology Data Exchange (ETDEWEB)

Park, Won Dong; Bahn, Chi Bum [Pusan National University, Busan (Korea, Republic of)

2015-10-15

ASTM G30 suggests that the applied strain can be calculated by dividing thickness by a bend radius. It should be noted, however, that the formula is reliable under an assumption that the ratio of thickness to bend radius is less than 0.2. Typically, to increase the applied stress/strain, the ratio of thickness to bend radius becomes larger than 0.2. This suggests that the estimated strain values by ASTM G30 are not reliable to predict the actual residual strain state of the highly deformed U-bend specimen. For this reason, finite element analysis (FEA) for the bending process of Ubend specimens was conducted by using a commercial finite element analysis software ABAQUS. ver.6.14- 2;2014. From the results of FEA, PWSCC initiation time and U-bend specimen size can be determined exactly. Since local stress and strain have a significant effect on the initiation of PWSCC, it was inappropriate to apply results of ASTM G30 to the PWSCC test directly. According to results of finite element analysis (FEA), elastic relaxation can cause inaccuracy in intended final residual stress. To modify this inaccuracy, additional process reducing the spring back is required. However this additional process also may cause uncertainty of stress/strain state. Therefore, the U-bending specimen size which is not creating uncertainty should be optimized and selected. With the bending radius of 8.3 mm, the thickness of 3 mm and the roller distance of 32.6 mm, calculated maximum stress and strain were 670 MPa and 0.21, respectively.

Study plan for conducting a section 316(a) demonstration: K-Reactor cooling tower, Savannah River Site

International Nuclear Information System (INIS)

Paller, M.H.

1991-02-01

The K Reactor at the Savannah River Site (SRS) began operation in 1954. The K-Reactor pumped secondary cooling water from the Savannah River and discharged directly to the Indian Grave Branch, a tributary of Pen Branch which flows to the Savannah River. During earlier operations, the temperature and discharge rates of cooling water from the K-reactor were up to approximately 70 degree C and 400 cfs, substantially altering the thermal and flow regimes of this stream. These discharges resulted in adverse impacts to the receiving stream and wetlands along the receiving stream. As a component of a Consent Order (84-4-W as amended) with the South Carolina Department of Health and Environmental Control, the Department of Energy (DOE) evaluated the alternatives for cooling thermal effluents from K Reactor and concluded that a natural draft recirculating cooling tower should be constructed. The cooling tower will mitigate thermal and flow factors that resulted in the previous impacts to the Indian Grave/Pen Branch ecosystem. The purpose of the proposed biological monitoring program is to provide information that will support a Section 316(a) Demonstration for Indian Grave Branch and Pen Branch when K-Reactor is operated with the recirculating cooling tower. The data will be used to determine that Indian Grave Branch and Pen Branch support Balanced Indigenous Communities when K-Reactor is operated with a recirculating cooling tower. 4 refs., 1 fig. 1 tab

Scaling analysis for a Savannah River reactor scaled model integral system

International Nuclear Information System (INIS)

Boucher, T.J.; Larson, T.K.; McCreery, G.E.; Anderson, J.L.

1990-11-01

801The Savannah River Laboratory has requested that the Idaho National Engineering Laboratory perform an analysis to help define, examine, and assess potential concepts for the design of a scaled integral hydraulics test facility representative of the current Savannah River Plant reactor design. In this report the thermal-hydraulic phenomena of importance (based on the knowledge and experience of the authors and the results of the joint INEL/TPG/SRL phenomena identification and ranking effort) to reactor safety during the design basis loss-of-coolant accident were examined and identified. Established scaling methodologies were used to develop potential concepts for integral hydraulic testing facilities. Analysis is conducted to examine the scaling of various phenomena in each of the selected concepts. Results generally support that a one-fourth (1/4) linear scale visual facility capable of operating at pressures up to 350 kPa (51 psia) and temperatures up to 330 K (134 degree F) will scale most hydraulic phenomena reasonably well. However, additional research will be necessary to determine the most appropriate method of simulating several of the reactor components, since the scaling methodology allows for several approaches which may only be assessed via appropriate research. 34 refs., 20 figs., 14 tabs

High-temperature reverse-bend fatigue strength of Inconel Alloy 625

International Nuclear Information System (INIS)

Purohit, A.; Greenfield, I.G.; Park, K.B.

1983-06-01

Inconel 625 has been selected as the clad material for Upgraded Transient Reactor Test Facility (TREAT Upgrade or TU) fuel assemblies. The range of temperatures investigated is 900 to 1100 0 C. A reverse-bend fatigue test program was selected as the most-effective method of determining the fatigue characteristics of Inconel alloy 625 sheet metal. The paper describes the reverse bend fatigue experiments, the results obtained, and the analysis of data

Flooding characteristics of gas-liquid two-phase flow in a horizontal U bend pipe

International Nuclear Information System (INIS)

Sakaguchi, T.; Hosokawa, S.; Fujii, Y.

1995-01-01

For next-generation nuclear reactors, hybrid safety systems which consist of active and passive safety systems have been planned. Steam generators with horizontal U bend pipelines will be used as one of the passive safety systems. It is required to clarify flow characteristics, especially the onset of flooding, in the horizontal U bend pipelines in order to examine their safety. Flooding in vertical pipes has been studied extensively. However, there is little study on flooding in the horizontal U bend pipelines. It is supposed that the onset of flooding in the horizontal U bend pipelines is different from that in vertical pipes. On the other hand, liquid is generated due to condensation of steam in pipes of the horizontal steam generators at the loss of coolant accident because the steam generators will be used as a condenser of a cooling system of steam from the reactor. It is necessary to simulate this situation by the supply of water at the middle of horizontal pipe. In the present paper, experiments were carried out using a horizontal U bend pipeline with a liquid supply section in the midway of pipeline. The onset of flooding in the horizontal U bend pipeline was measured. Effects of the length of horizontal pipe and the radius of U bend on the onset of flooding were discussed

Flooding characteristics of gas-liquid two-phase flow in a horizontal U bend pipe

Energy Technology Data Exchange (ETDEWEB)

Sakaguchi, T.; Hosokawa, S.; Fujii, Y. [Kobe Univ. (Japan)] [and others

1995-09-01

For next-generation nuclear reactors, hybrid safety systems which consist of active and passive safety systems have been planned. Steam generators with horizontal U bend pipelines will be used as one of the passive safety systems. It is required to clarify flow characteristics, especially the onset of flooding, in the horizontal U bend pipelines in order to examine their safety. Flooding in vertical pipes has been studied extensively. However, there is little study on flooding in the horizontal U bend pipelines. It is supposed that the onset of flooding in the horizontal U bend pipelines is different from that in vertical pipes. On the other hand, liquid is generated due to condensation of steam in pipes of the horizontal steam generators at the loss of coolant accident because the steam generators will be used as a condenser of a cooling system of steam from the reactor. It is necessary to simulate this situation by the supply of water at the middle of horizontal pipe. In the present paper, experiments were carried out using a horizontal U bend pipeline with a liquid supply section in the midway of pipeline. The onset of flooding in the horizontal U bend pipeline was measured. Effects of the length of horizontal pipe and the radius of U bend on the onset of flooding were discussed.

Service water chemical cleaning at River Bend gets results

International Nuclear Information System (INIS)

Brice, T.O.; Glover, W.A.

1994-01-01

The largest known Service Water System (SWS) chemical cleaning ever performed at a nuclear plant was successfully completed at, River Bend Station. Corrosion product buildup was observed during system inspections in the first operating cycle and the first refueling outage in 1987. Under deposit corrosion was followed with microbiologically influenced corrosion (MIC) occurring as a later stage under deposits. The heavy corrosion caused blockage of heat exchanger tubes, fouling of valve seats, and general flow blockage throughout the system. Various options were evaluated for restoring the SWS back to an acceptable long term operating condition. The large scale chemical cleaning performed arrested the corrosion by removing the deposits down to the bare metal surfaces and leaving behind a protective passivation layer. After the cleaning, the open recirculating SWS was converted to a closed system. The implementation of a molybdate/nitrate water treatment program with a copper corrosion inhibitor maintained at a high pH (8.5--10.5) has significantly reduced corrosion rates in the closed system. This should extend the life of the SWS piping for the remaining life of the plant. Several field tests were conducted to qualify the process and demonstrate its ability to achieve acceptable cleaning results prior to being used on a larger scale. In the summer of 1992, temporary and permanent modifications were installed to divide the SWS into two separate cleaning loops for the system wide cleaning. The SWS chemical was successfully performed and completed on schedule during the fourth refueling outage. Post cleaning inspections at various locations throughout the Service Water System showed the process to be very effective at complete deposit removal

Evaluation of nuclear facility decommissioning projects. Project summary report, Elk River Reactor

International Nuclear Information System (INIS)

Miller, R.L.; Adams, J.A.

1982-12-01

This report summarizes information concerning the decommissioning of the Elk River Reactor. Decommissioning data from available documents were input into a computerized data-handling system in a manner that permits specific information to be readily retrieved. The information is in a form that assists the Nuclear Regulatory Commission in its assessment of decommissioning alternatives and ALARA methods for future decommissionings projects. Samples of computer reports are included in the report. Decommissioning of other reactors, including NRC reference decommissioning studies, will be described in similar reports

SAVANNAH RIVER SITE R REACTOR DISASSEMBLY BASIN IN SITU DECOMMISSIONING

Energy Technology Data Exchange (ETDEWEB)

Langton, C.; Blankenship, J.; Griffin, W.; Serrato, M.

2009-12-03

The US DOE concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate in tact, structurally sound facilities that are no longer needed for their original purpose of, i.e., generating (reactor facilities), processing(isotope separation facilities) or storing radioactive materials. The 105-R Disassembly Basin is the first SRS reactor facility to undergo the in-situ decommissioning (ISD) process. This ISD process complies with the105-R Disassembly Basin project strategy as outlined in the Engineering Evaluation/Cost Analysis for the Grouting of the R-Reactor Disassembly Basin at the Savannah River Site and includes: (1) Managing residual water by solidification in-place or evaporation at another facility; (2) Filling the below grade portion of the basin with cementitious materials to physically stabilize the basin and prevent collapse of the final cap - Sludge and debris in the bottom few feet of the basin will be encapsulated between the basin floor and overlying fill material to isolate if from the environment; (3) Demolishing the above grade portion of the structure and relocating the resulting debris to another location or disposing of the debris in-place; and (4) Capping the basin area with a concrete slab which is part of an engineered cap to prevent inadvertent intrusion. The estimated total grout volume to fill the 105-R Reactor Disassembly Basin is 24,424 cubic meters or 31,945 cubic yards. Portland cement-based structural fill materials were design and tested for the reactor ISD project and a placement strategy for stabilizing the basin was developed. Based on structural engineering analyses and work flow considerations, the recommended maximum lift height is 5 feet with 24 hours between lifts. Pertinent data and information related to the SRS 105-R-Reactor Disassembly Basin in-situ decommissioning include: regulatory documentation, residual water management, area preparation activities, technology needs, fill material designs

Dynamic shear-bending buckling experiments of cylindrical shells

International Nuclear Information System (INIS)

Hagiwara, Y.; Akiyama, H.

1995-01-01

Dynamic experimental studies of the plastic shear/bending buckling of cylindrical shells were performed. They clarified the inelastic response reduction and the seismic margin of FBR reactor vessels. The test results were incorporated into the draft of the seismic buckling design guidelines of FBR. (author). 15 refs., 3 figs

Loss-of-coolant accident analysis of the Savannah River new production reactor design

International Nuclear Information System (INIS)

Maloney, K.J.; Pryor, R.J.

1990-11-01

This document contains the loss-of-coolant accident analysis of the representative design for the Savannah River heavy water new production reactor. Included in this document are descriptions of the primary system, reactor vessel, and loss-of-coolant accident computer input models, the results of the cold leg and hot leg loss-of-coolant accident analyses, and the results of sensitivity calculations for the cold leg loss-of-coolant accident. 5 refs., 50 figs., 4 tabs

Inundation and draining of the Trinity River floodplain associated with extreme precipitation from Hurricane Harvey, east Texas, USA

Science.gov (United States)

Hassenruck-Gudipati, H. J.; Goudge, T. A.; Mohrig, D. C.

2017-12-01

Rivers swelled up beyond their historic high-water marks due to precipitation from Hurricane Harvey. We used Harvey-induced flooding to investigate the flow connectivity between the coastal Trinity River and its floodplain by measuring water depth and velocity, as well as sediment-transporting conditions on the natural levee that separates the two. River discharge within the study area peaked at a historic high of 3600 cubic meters per second on September 1, 2017. The levees on two river bends were investigated on September 3 and 4 in order to characterize the hydraulic connectivity between the channel and its floodplain during the early falling limb of this flood. On September 3, a river bend located approximately 28km upstream of the river mouth was visited. Water was overtopping the levee crest at this location, 30m away from the levee crest. This overland flow only experienced about a threefold reduction in average velocity to 0.16 m/s (in 2.2 m of water) 600m away from the levee crest. On September 4, a river bend approximately 59km upstream of the river mouth was investigated. Even though the river stage was at the National Weather Service major flood stage, the levee crest separating the river and floodplain was emergent. Regardless of this local disconnect between the river and its floodplain, substantial and variable drainage velocities were measured depending on drainage patterns controlled by local topography. Velocities measured in shallow water immediately adjacent to the emergent levee were low (< 0.05 m/s in 0.2 m of water). The highest drainage velocity ( 0.18 m/s in 1.7 m of water) associated with the upstream river-bend was measured at 750m from the channel and was similar in magnitude to those recorded for the distal inundating overland flow a day before on the downstream river-bend. Results from this work highlight the maintenance of high flow velocities across the distal floodplain even during its drainage phase. The transport of sediment

Bar and channel evolution in meandering and braiding rivers using physics-based modeling

NARCIS (Netherlands)

Schuurman, F.

2015-01-01

Rivers are among the most dynamic earth surface systems. Some rivers meander, forming bends that migrate, reshape and have inner-bend bars. Other rivers form a complicated braided pattern of branches, islands and mid-channel bars. Thorough understanding of their morphodynamics is important for

Dynamic simulation of the 2 MWt slowpoke heating reactor

International Nuclear Information System (INIS)

Tseng, C.M.; Lepp, R.M.

1982-04-01

A 2 MWt SLOWPOKE reactor, intended for commercial space heating, is being developed at the Chalk River Nuclear Laboratories. A small-signal dynamic simulation of this reactor, without closed-loop control, was developed. Basic equations were used to describe the physical phenomena in each kf the eight reactor subsystems. These equations were then linearized about the normal operation conditions and rearranged in a dimensionless form for implementation. The overall simulation is non-linear. Slow transient responses (minutes to days) of the simulation to both reactivity and temperature perturbations were measured at full power. In all cases the system reached a new steady state in times varying from 12 h to 250 h. These results illustrate the benefits of the inherent negative reactivity feedback of this reactor concept. The addition of closed-loop control using core outlet temperature as the controlled variable to move a beryllium reflector is also examined

Microstructure in Zircaloy Creep Tested in the R2 Reactor

International Nuclear Information System (INIS)

Pettersson, Kjell

2004-12-01

Tubular specimens of Zircaloy-4 have been creep tested in bending in the R2 reactor in Studsvik. The creep deformation in the reactor core is accelerated in comparison with creep deformation outside the reactor core. The possible mechanisms behind this behaviour are described briefly. In order to determine which the actual mechanism is, the microstructure of the material creep tested in the R2 reactor has been examined by transmission electron microscopy. Due to the bending, material subjected to both tensile and compressive stress during creep was available. Since some of the proposed mechanisms might give microstructures which are different when the material is subjected to compressive or tensile stress it was assumed that examination of both types of material would give valuable information with regard to the operating mechanism. The result of the examination was that in the as-irradiated condition there were no obvious differences detected between materials which had been deformed in tension or compression. After a heat treatment to coarsen the irradiation induced microstructure there were still no significant differences between the two types of material. However it was now observed that in addition to dislocation loops the microstructure also contained network dislocations which presumably had been invisible in the electron microscope before heat treatment due to the high density of small dislocation loops in this state. It is therefore concluded that the most probable mechanism for irradiation creep in this case is climb and glide of the network dislocations. The role of irradiation is two-fold: It accelerates climb due to the production of point defects of which more interstitials than vacancies arrive to the network dislocations stopped at an obstacles. This leads to a net climb after which a dislocation is released from the obstacle and an amount of glide takes place. The second effect is the production of loops which serve as an increasing density of

A bend thickness sensitivity study of Candu feeder piping

International Nuclear Information System (INIS)

Li, M.; Aggarwal, M.L.; Meysner, A.; Micelotta, C.

2005-01-01

In CANDU reactors, feeder bends close to the connection at the fuel channel may be subjected to the highest Flow Accelerated Corrosion (FAC) and stresses. Feeder pipe stress analysis is crucial in the life extension of aging CANDU plants. Typical feeder pipes are interconnected by upper link plates and spacers. It is well known that the stresses at the bends are sensitive to the local bend thicknesses. It is also known from the authors' study (Li and et al, 2005) that feeder inter linkage effect is significant and cannot be ignored. The field measurement of feeder bend thickness is difficult and may be subjected to uncertainty in accuracy. Hence, it is desirable to know how the stress on a subject feeder could be affected by the bend thickness variation of the neighboring feeders. This effect cannot be evaluated by the traditional 'single' feeder model approach. In this paper, the 'row' and 'combined' models developed in the previous study (Li and et al, 2005), which include the feeder interactions, are used to investigate the sensitivity of bend thickness. A series of random thickness bounded by maximum and minimum measured values were applied to feeders in the model. The results show that an individual feeder is not sensitive to the bend thickness variation of the remaining feeders in the model, but depends primarily on its own bend thickness. The highest stress at a feeder always occurs when the feeder has the smallest possible bend thickness. A minimum acceptable bend thickness for individual feeders can be computed by an iterative computing process. The dependency of field thickness measurement and the amount of required analysis work can be greatly reduced. (authors)

Structural integrity of water reactor pressure boundary components

International Nuclear Information System (INIS)

Loss, F.J.

1977-01-01

The dynamic fracture toughness was determined as a function of temperature for three-point bend specimens of A533-B, A508-2, and A302-B steels. Crack propagation rates at 288 0 C in a water reactor environment were determined for A533-B and A508-2. Radiation-induced degradation of notch toughness of reactor steels and welds was explored. The ''warm prestress'' occurring in a flawed reactor vessel following a LOCA and operation of ECCS was studied. 25 figures

Core melt progression and consequence analysis methodology development in support of the Savannah River Reactor PSA

International Nuclear Information System (INIS)

O'Kula, K.R.; Sharp, D.A.; Amos, C.N.; Wagner, K.C.; Bradley, D.R.

1992-01-01

A three-level Probabilistic Safety Assessment (PSA) of production reactor operation has been underway since 1985 at the US Department of Energy's Savannah River Site (SRS). The goals of this analysis are to: Analyze existing margins of safety provided by the heavy-water reactor (HWR) design challenged by postulated severe accidents; Compare measures of risk to the general public and onsite workers to guideline values, as well as to those posed by commercial reactor operation; and Develop the methodology and database necessary to prioritize improvements to engineering safety systems and components, operator training, and engineering projects that contribute significantly to improving plant safety. PSA technical staff from the Westinghouse Savannah River Company (WSRC) and Science Applications International Corporation (SAIC) have performed the assessment despite two obstacles: A variable baseline plant configuration and power level; and a lack of technically applicable code methodology to model the SRS reactor conditions. This paper discusses the detailed effort necessary to modify the requisite codes before accident analysis insights for the risk assessment were obtained

Review of Savannah River Site K Reactor inservice inspection and testing restart program

International Nuclear Information System (INIS)

Anderson, M.T.; Hartley, R.S.; Kido, C.

1992-09-01

Inservice inspection (ISI) and inservice testing (IST) programs are used at commercial nuclear power plants to monitor the pressure boundary integrity and operability of components in important safety-related systems. The Department of Energy (DOE) - Office of Defense Programs (DP) operates a Category A (> 20 MW thermal) production reactor at the Savannah River Site (SRS). This report represents an evaluation of the ISI and IST practices proposed for restart of SRS K Reactor as compared, where applicable, to current ISI/IST activities of commercial nuclear power facilities

Alignment Compensation for Bending Radius in TI 2 Transfer Line Magnets

CERN Document Server

Weterings, W

2004-01-01

The optics file for the TI 2 transfer lines specifies the position of the bending magnets assuming that the beam enters and exists at the centre of the vacuum pipe. In order to disbribute the deflected beam evenly inside the vacuum tube, the alignment has to be compensated by moving the magnets half of the beam deflection away from the centre of the bending radius. In this note the saggitas of the various TI 2 magnets are calculated and the alignment displacements tabulated for future reference.

Axial power monitoring uncertainty in the Savannah River Reactors

International Nuclear Information System (INIS)

Losey, D.C.; Revolinski, S.M.

1990-01-01

The results of this analysis quantified the uncertainty associated with monitoring the Axial Power Shape (APS) in the Savannah River Reactors. Thermocouples at each assembly flow exit map the radial power distribution and are the primary means of monitoring power in these reactors. The remaining uncertainty in power monitoring is associated with the relative axial power distribution. The APS is monitored by seven sensors that respond to power on each of nine vertical Axial Power Monitor (APM) rods. Computation of the APS uncertainty, for the reactor power limits analysis, started with a large database of APM rod measurements spanning several years of reactor operation. A computer algorithm was used to randomly select a sample of APSs which were input to a code. This code modeled the thermal-hydraulic performance of a single fuel assembly during a design basis Loss-of Coolant Accident. The assembly power limit at Onset of Significant Voiding was computed for each APS. The output was a distribution of expected assembly power limits that was adjusted to account for the biases caused by instrumentation error and by measuring 7 points rather than a continuous APS. Statistical analysis of the final assembly power limit distribution showed that reducing reactor power by approximately 3% was sufficient to account for APS variation. This data confirmed expectations that the assembly exit thermocouples provide all information needed for monitoring core power. The computational analysis results also quantified the contribution to power limits of the various uncertainties such as instrumentation error

SAVANNAH RIVER SITE R-REACTOR DISASSEMBLY BASIN IN-SITU DECOMMISSIONING -10499

Energy Technology Data Exchange (ETDEWEB)

Langton, C.; Serrato, M.; Blankenship, J.; Griffin, W.

2010-01-04

The US DOE concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate intact, structurally sound facilities that are no longer needed for their original purpose, i.e., generating (reactor facilities), processing(isotope separation facilities) or storing radioactive materials. The 105-R Disassembly Basin is the first SRS reactor facility to undergo the in-situ decommissioning (ISD) process. This ISD process complies with the 105-R Disassembly Basin project strategy as outlined in the Engineering Evaluation/Cost Analysis for the Grouting of the R-Reactor Disassembly Basin at the Savannah River Site and includes: (1) Managing residual water by solidification in-place or evaporation at another facility; (2) Filling the below grade portion of the basin with cementitious materials to physically stabilize the basin and prevent collapse of the final cap - Sludge and debris in the bottom few feet of the basin will be encapsulated between the basin floor and overlying fill material to isolate it from the environment; (3) Demolishing the above grade portion of the structure and relocating the resulting debris to another location or disposing of the debris in-place; and (4) Capping the basin area with a concrete slab which is part of an engineered cap to prevent inadvertent intrusion. The estimated total grout volume to fill the 105-R Reactor Disassembly Basin is 24,384 cubic meters or 31,894 cubic yards. Portland cement-based structural fill materials were designed and tested for the reactor ISD project, and a placement strategy for stabilizing the basin was developed. Based on structural engineering analyses and material flow considerations, maximum lift heights and differential height requirements were determined. Pertinent data and information related to the SRS 105-R Reactor Disassembly Basin in-situ decommissioning include: regulatory documentation, residual water management, area preparation activities, technology needs, fill material

Lessons learned from the licensing process for the Clinch River Breeder Reactor Plant

International Nuclear Information System (INIS)

Dickson, P.W.; Clare, G.H.

1991-01-01

This paper presents the experience of licensing a specific liquid-metal fast breeder reactor (LMFBR), the Clinch River Breader Reactor Plant (CRBRP). It was a success story in that the licensing process was accomplished in a very short time span. The actions of the applicant and the actions of the US Nuclear Regulatory Commission (NRC) in response are presented and discussed to provide guidance to future efforts to license unconventional reactors. The history is told from the perspective of the authors. As such, some of the reasons given for success or lack of success are subjective interpretations. Nevertheless, the authors' positions provided them an excellent viewpoint to make these judgements. During the second phase of the licensing process, they were the CRBRP Technical Director and the Licensing Manager, respectively, for the Westinghouse Electric Corporation, the prime contractor for the reactor plant

Clinch River Breeder Reactor secondary control rod system

International Nuclear Information System (INIS)

McKeehan, E.R.; Sim, R.G.

1977-01-01

The shutdown system for the Clinch River Breeder Reactor (CRBR) includes two independent systems--a primary and a secondary system. The Secondary Control Rod System (SCRS) is a new design which is being developed by General Electric to be independent from the primary system in order to improve overall shutdown reliability by eliminating potential common-mode failures. The paper describes the status of the SCRS design and fabrication and testing activities. Design verification testing on the component level is largely complete. These component tests are covered with emphasis on design impact results. A prototype unit has been manufactured and system level tests in sodium have been initiated

Processing Tritiated Water at the Savannah River Site: A Production-Scale Demonstration of a palladium membrane reactor

International Nuclear Information System (INIS)

Sessions, K

2004-01-01

The Palladium Membrane Reactor (PMR) process was installed in the Tritium Facilities at the Savannah River Site to perform a production-scale demonstration for the recovery of tritium from tritiated water adsorbed on molecular sieve (zeolite). Unlike the current recovery process that utilizes magnesium, the PMR offers a means to process tritiated water in a more cost effective and environmentally friendly manner. The design and installation of the large-scale PMR process was part of a collaborative effort between the Savannah River Site and Los Alamos National Laboratory. The PMR process operated at the Savannah River Site between May 2001 and April 2003. During the initial phase of operation the PMR processed thirty-four kilograms of tritiated water from the Princeton Plasma Physics Laboratory. The water was processed in fifteen separate batches to yield approximately 34,400 liters (STP) of hydrogen isotopes. Each batch consisted of round-the-clock operations for approximately nine days. In April 2003 the reactor's palladium-silver membrane ruptured resulting in the shutdown of the PMR process. Reactor performance, process performance and operating experiences have been evaluated and documented. A performance comparison between PMR and current magnesium process is also documented

Fracture assessment of Savannah River Reactor carbon steel piping

International Nuclear Information System (INIS)

Mertz, G.E.; Stoner, K.J.; Caskey, G.R.; Begley, J.A.

1991-01-01

The Savannah River Site (SRS) production reactors have been in operation since the mid-1950's. One postulated failure mechanism for the reactor piping is brittle fracture of the original A285 and A53 carbon steel piping. Material testing of archival piping determined (1) the static and dynamic tensile properties; (2) Charpy impact toughness; and (3) the static and dynamic compact tension fracture toughness properties. The nil-ductility transition temperature (NDTT), determined by Charpy impact test, is above the minimum operating temperature for some of the piping materials. A fracture assessment was performed to demonstrate that potential flaws are stable under upset loading conditions and minimum operating temperatures. A review of potential degradation mechanisms and plant operating history identified weld defects as the most likely crack initiation site for brittle fracture. Piping weld defects, as characterized by radiographic and metallographic examination, and low fracture toughness material properties were postulated at high stress locations in the piping. Normal operating loads, upset loads, and residual stresses were assumed to act on the postulated flaws. Calculated allowable flaw lengths exceed the size of observed weld defects, indicating adequate margins of safety against brittle fracture. Thus, a detailed fracture assessment was able to demonstrate that the piping systems will not fail by brittle fracture, even though the NDTT for some of the piping is above the minimum system operating temperature

Phipps Bend Nuclear Energy Project. Community impact assessment. Final report

International Nuclear Information System (INIS)

Snapp, P.C.; Teilhet, A.; Newsom, R.; Bond, M.; Garland, M.

1977-01-01

In late 1977, the Tennessee Valley Authority (TVA) proposed to build a 2 unit nuclear plant at Phipps Bend on the Holston River east of Surgoinsville, Tennessee. Total estimated cost is 1.6 billion dollars, with a generating capacity of 2,600,000 kilowatts. The facility will have an impact on Hawkins, Greene and Sullivan counties with 2,500 construction employees, a permanent work force of 300, increased availability of energy to stimulate new capital investment and the local government will need to deal with these. This report analyzed the facilities of each community in the impacted area and recommended certain action for infrastructure acquisition or improvements

Methods for Quantifying Shallow-Water Habitat Availability in the Missouri River

Energy Technology Data Exchange (ETDEWEB)

Hanrahan, Timothy P.; Larson, Kyle B.

2012-04-09

As part of regulatory requirements for shallow-water habitat (SWH) restoration, the U.S. Army Corps of Engineers (USACE) completes periodic estimates of the quantity of SWH available throughout the lower 752 mi of the Missouri River. To date, these estimates have been made by various methods that consider only the water depth criterion for SWH. The USACE has completed estimates of SWH availability based on both depth and velocity criteria at four river bends (hereafter called reference bends), encompassing approximately 8 river miles within the lower 752 mi of the Missouri River. These estimates were made from the results of hydraulic modeling of water depth and velocity throughout each bend. Hydraulic modeling of additional river bends is not expected to be completed for deriving estimates of available SWH. Instead, future estimates of SWH will be based on the water depth criterion. The objective of this project, conducted by the Pacific Northwest National Laboratory for the USACE Omaha District, was to develop geographic information system methods for estimating the quantity of available SWH based on water depth only. Knowing that only a limited amount of water depth and channel geometry data would be available for all the remaining bends within the lower 752 mi of the Missouri River, the intent was to determine what information, if any, from the four reference bends could be used to develop methods for estimating SWH at the remaining bends. Specifically, we examined the relationship between cross-section channel morphology and relative differences between SWH estimates based on combined depth and velocity criteria and the depth-only criterion to determine if a correction factor could be applied to estimates of SWH based on the depth-only criterion. In developing these methods, we also explored the applicability of two commonly used geographic information system interpolation methods (TIN and ANUDEM) for estimating SWH using four different elevation data

Structural analysis of the Upper Internals Structure for the Clinch River Breeder Reactor Plant

International Nuclear Information System (INIS)

Houtman, J.L.

1979-01-01

The Upper Internals Structure (UIS) of the Clinch River Breeder Reactor Plant (CRBRP) provides control of core outlet flow to prevent severe thermal transients from occuring at the reactor vessel and primary heat transport outlet piping, provides instrumentation to monitor core performance, provides support for the control rod drivelines, and provides secondary holddown of the core. All of the structural analysis aspects of assuring the UIS is structurally adequate are presented including simplified and rigorous inelastic analysis methods, elevated temperature criteria, environmental effects on material properties, design techniques, and manufacturing constraints

Steady-state and loss-of-pumping accident analyses of the Savannah River new production reactor representative design

International Nuclear Information System (INIS)

Pryor, R.J.; Maloney, K.J.

1990-10-01

This document contains the steady-state and loss-of-pumping accident analysis of the representative design for the Savannah River heavy water new production reactor. A description of the reactor system and computer input model, the results of the steady-state analysis, and the results of four loss-of-pumping accident calculations are presented. 5 refs., 37 figs., 4 tabs

Safety-Evaluation Report related to the construction of the Clinch River Breeder Reactor Plant. Docket No. 50-537

International Nuclear Information System (INIS)

1983-03-01

The Safety-Evaluation Report for the application by the United States Department of Energy, Tennessee Valley Authority, and the Project Management Corporation, as applicants and owners, for a license to construct the Clinch River Breeder Reactor Plant (docket No. 50-537) has been prepared by the Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Commission. The facility will be located on the Clinch River approximately 12 miles southwest of downtown Oak Ridge and 25 miles west of Knoxville, Tennessee. Subject to resolution of the items discussed in this report, the staff concludes that the construction permit requested by the applicants should be issued

Magnetoelastic bending and snapping of ferromagnetic plates in oblique magnetic fields

International Nuclear Information System (INIS)

Zhou Youhe

1995-01-01

Ferritic stainless steel has been considered for structural components such as first walls and blankets of fusion power reactors because the material shows low rates of irradiation swelling. Since it is magnetizable, the magnetoelastic interaction between magnetic field and deformation of the structures in a fusion reactor is so strong that their safety is of concern due to the magnetoelastic bending, buckling and magnetic damping, etc. Basic research of the magnetoelastic characteristics of ferromagnetic plate has been paid special attention by researchers. In this paper, the magnetoelastic bending and snapping are studied for a ferromagnetic plate in an oblique magnetic field. The theoretical model is based on the variational principle where the functional is employed as real total energy in the system including external work. The obtained expression of magnetic force on the plate is the same as that derived from the dipole model when the total magnetic field in the ferromagnetic medium is considered. In order to effectively solve the nonlinearly coupled interaction problem between magnetic field and mechanical deformation, a numerical program combining the finite element method for analyzing the magnetic field with the finite difference technique for finding out the bending deformation of the plate is employed to obtain the solution of magnetoelastic bending of a soft ferromagnetic plate. The numerical calculations are carried out for the typical example of a ferromagnetic cantilevered beam-plate in an oblique magnetic field. From the bending curves, that is the tip deflection versus applied magnetic fields, the critical magnetic field for the magnetoelastic snapping is predicted by the Southwell plot. The theoretical predictions show that the critical magnetic field decreases with the increase in incident angle of the oblique magnetic field. By the effect of incident angle on the magnetic buckling, the discrepancy between theoretical and experimental data can

Weibull statistical analysis of Krouse type bending fatigue of nuclear materials

Energy Technology Data Exchange (ETDEWEB)

Haidyrah, Ahmed S., E-mail: ashdz2@mst.edu [Nuclear Engineering, Missouri University of Science & Technology, 301 W. 14th, Rolla, MO 65409 (United States); Nuclear Science Research Institute, King Abdulaziz City for Science and Technology (KACST), P.O. Box 6086, Riyadh 11442 (Saudi Arabia); Newkirk, Joseph W. [Materials Science & Engineering, Missouri University of Science & Technology, 1440 N. Bishop Ave, Rolla, MO 65409 (United States); Castaño, Carlos H. [Nuclear Engineering, Missouri University of Science & Technology, 301 W. 14th, Rolla, MO 65409 (United States)

2016-03-15

A bending fatigue mini-specimen (Krouse-type) was used to study the fatigue properties of nuclear materials. The objective of this paper is to study fatigue for Grade 91 ferritic-martensitic steel using a mini-specimen (Krouse-type) suitable for reactor irradiation studies. These mini-specimens are similar in design (but smaller) to those described in the ASTM B593 standard. The mini specimen was machined by waterjet and tested as-received. The bending fatigue machine was modified to test the mini-specimen with a specially designed adapter. The cycle bending fatigue behavior of Grade 91 was studied under constant deflection. The S–N curve was created and mean fatigue life was analyzed using mean fatigue life. In this study, the Weibull function was predicted probably for high stress to low stress at 563, 310 and 265 MPa. The commercial software Minitab 17 was used to calculate the distribution of fatigue life under different stress levels. We have used 2 and 3- parameters Weibull analysis to introduce the probability of failure. The plots indicated that the 3- parameter Weibull distribution fits the data well.

Weibull statistical analysis of Krouse type bending fatigue of nuclear materials

International Nuclear Information System (INIS)

Haidyrah, Ahmed S.; Newkirk, Joseph W.; Castaño, Carlos H.

2016-01-01

A bending fatigue mini-specimen (Krouse-type) was used to study the fatigue properties of nuclear materials. The objective of this paper is to study fatigue for Grade 91 ferritic-martensitic steel using a mini-specimen (Krouse-type) suitable for reactor irradiation studies. These mini-specimens are similar in design (but smaller) to those described in the ASTM B593 standard. The mini specimen was machined by waterjet and tested as-received. The bending fatigue machine was modified to test the mini-specimen with a specially designed adapter. The cycle bending fatigue behavior of Grade 91 was studied under constant deflection. The S–N curve was created and mean fatigue life was analyzed using mean fatigue life. In this study, the Weibull function was predicted probably for high stress to low stress at 563, 310 and 265 MPa. The commercial software Minitab 17 was used to calculate the distribution of fatigue life under different stress levels. We have used 2 and 3- parameters Weibull analysis to introduce the probability of failure. The plots indicated that the 3- parameter Weibull distribution fits the data well.

Thoria-fuel irradiation. Program to irradiate 80% ThO2/20% UO2 ceramic pellets at the Savannah River Plant

International Nuclear Information System (INIS)

Pickett, J.B.

1982-02-01

This report describes the fabrication of proliferation-resistant thorium oxide/uranium oxide ceramic fuel pellets and preparations at the Savannah River Laboratory (SRL) to irradiate those materials. The materials were fabricated in order to study head end process steps (decladding, tritium removal, and dissolution) which would be required for an irradiated proliferation-resistant thorium based fuel. The thorium based materials were also to be studied to determine their ability to withstand average commercial light water reactor (LWR) irradiation conditions. This program was a portion of the Thorium Fuel Cycle Technology (TFCT) Program, and was coordinated by the Oak Ridge National Laboratory (ORNL) under the Consolidated Fuel Reprocessing Program (CFRP). The fuel materials were to be irradiated in a Savannah River Plant (SRP) reactor at conditions simulating the heat ratings and burnup of a commercial LWR. The program was terminated due to a de-emphasis of the TFCT Program, following completion of the fabrication of the fuel and the modified assemblies which were to be used in the SRP reactor. The reactor grade ceramic pellets were fabricated for SRL by Battelle, Pacific Northwest Laboratories. Five fuel types were prepared: 100% UO 2 pellets (control); 80% ThO 2 /20% UO 2 pellets; approximately 80% ThO 2 /20% UO 2 + 0.25 CaO (dissolution aid) pellets; 100% UO 2 hybrid pellets (prepared from sol-gel microspheres); and 100% ThO 2 pellets (control). All of the fuel materials were transferred to SRL from PNL and were stored pending a subsequent reactivation of the TFCT Programs

Stress analysis of feeder bends using neutrons: new results and cumulative impacts

Energy Technology Data Exchange (ETDEWEB)

Banks, D.; Donaberger, R. [Canadian Neutron Beam Centre, Chalk River, ON (Canada); Leitch, B. [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Rogge, R.B. [Canadian Neutron Beam Centre, Chalk River, ON (Canada)

2014-07-01

Neutron diffraction has played a vital role in stress analysis of bends in carbon steel pipes, known as feeder pipes, in CANDU reactors. Due to incidents of cracking of feeders, extensive R&D programs to manage feeder cracking have been implemented over about ten years. We review the cumulative impacts of this research from the view point of the stress analysis using neutrons, and present new results by examining a feeder bend with a partial crack both experimentally using neutron diffraction and theoretically using a finite element model. (author)

Tritium sample analyses in the Savannah River and associated waterways following the K-reactor release of December 1991

International Nuclear Information System (INIS)

Beals, D.M.; Dunn, D.L.; Hall, G.; Kantelo, M.V.

1992-01-01

An unplanned release of tritiated water occurred at K reactor on SRS between 22-December and 25-December 1991. This water moved down through the effluent canal, Pen Branch, Steel Creek and finally to the Savannah River. Samples were collected in the Savannah River and associated waterways over a period of a month. The Environmental Technology Section (ETS) of the Savannah River Laboratory performed liquid scintillation analyses to monitor the passage of the tritiated water from SRS to the Atlantic Ocean

Flow Patterns and Morphological Changes in a Sandy Meander Bend during a Flood—Spatially and Temporally Intensive ADCP Measurement Approach

Directory of Open Access Journals (Sweden)

Elina Kasvi

2017-02-01

Full Text Available The fluvio-geomorphological processes in meander bends are spatially uneven in distribution. Typically, higher velocities and erosion take place near the outer bank beyond the bend apex, while the inner bend point bar grows laterally towards the outer bank, increasing the bend amplitude. These dynamics maintain the meander evolution. Even though this development is found in meandering rivers independent of soil or environmental characteristics, each river still seems to behave unpredictably. The special mechanisms that determine the rate and occasion of morphological changes remain unclear. The aim of this study is to offer new insights regarding flow-induced morphological changes in meander using a novel study approach. We focused on short-term and small-spatial-scale changes by conducting a spatially and temporally (daily intensive survey during a flood (a period of nine days with an ADCP attached to a remotely controlled mini-boat. Based on our analysis, the flood duration and the rate of discharge increase and decrease seems to play key roles in determining channel changes by controlling the flow velocities and depth and the backwater effect may have notable influence on the morphological processes. We discuss themes such as the interaction of inner and outer bend processes and the longer-term development of meander bends.

Safety Evaluation Report Restart of K-Reactor Savannah River Site

International Nuclear Information System (INIS)

1991-10-01

In April 1991, the Department of Energy (DOE) issued DOE/DP-0084T, ''Safety Evaluation Report Restart of K-Reactor Savannah River Site.'' The Safety Evaluation Report (SER) documents the results of DOE reviews and evaluations of the programmatic aspects of a large number of issues necessary to be satisfactorily addressed before restart. The issues were evaluated for compliance with the restart criteria included in the SER. The results of those evaluations determined that the restart criteria had been satisfied for some of the issues. However, for most of the issues at least part of the applicable restart criteria had not been found to be satisfied at the time the evaluations were prepared. For those issues, open or confirmatory items were identified that required resolution. In August 1991, DOE issued DOE/DP-0090T, ''Safety Evaluation Report Restart of K-Reactor Savannah River Site Supplement 1.'' That document was the first Supplement to the April 1991 SER, and documented the resolution of 62 of the open items identified in the SER. This document is the second Supplement to the April 1991 SER. This second SER Supplement documents the resolution of additional open times identified in the SER, and includes a complete list of all remaining SER open items. The resolution of those remaining open items will be documented in future SER Supplements. Resolution of all open items for an issue indicates that its associated restart criteria have been satisfied, and that DOE concludes that the programmatic aspects of the issue have been satisfactorily addressed

Device for removing a spent reactor core instrument tube

International Nuclear Information System (INIS)

Watanabe, Shigeru; Tsuji, Teruaki.

1980-01-01

Purpose: To easily and exactly execute works for removing a used reactor core instrument tube to be mounted in a reactor core from the lattice space of the core or for charging the tube into the lattice of the core. Constitution: When fuel assembly is pulled out of a reactor core and a spent reactor core instrument tube is then bent and removed from the core at periodical inspection time, a lower gripping unit integral with an upper gripping unit and a bending unit is provided at the lower end of a hanging rope of a winch, and lowered to the reactor core. Then, the spent reactor core instrument tube is gripped by the upper and lower gripping units, the bending unit is operated, the spent reactor core instrument tube is bent, and the tube is then pulled upwardly by the winch to remove the tube. (Aizawa, K.)

Bend me, shape me

CERN Multimedia

2002-01-01

A Japanese team has found a way to bend and shape silicon substrates by growing a thin layer of diamond on top. The technique has been proposed as an alternative to mechanical bending, which is currently used to make reflective lenses for X-ray systems and particle physics systems (2 paragraphs).

Methodology for definition of bending radius and pullback force in HDD (Horizontal Directional Drilling) operations

Energy Technology Data Exchange (ETDEWEB)

Silva, Danilo Machado L. da; Rodrigues, Marcos V. [Det Norske Veritas (DNV), Rio de Janeiro, RJ (Brazil); Venaas, Asle [Det Norske Veritas (DNV), Oslo (Norway); Medeiros, Antonio Roberto de [Subsea 7 (Brazil)

2009-12-19

Bending is a primary loading experienced by pipelines during installation and operation. Significant bending in the presence of tension is experienced during installation by the S-lay method, as the pipe conforms to the curvature of the stinger and beyond in the over bend region. Bending in the presence of external pressure is experienced in the sag bend of all major installation methods (e.g., reeling, J-lay, S-lay) as well as in free-spans on the sea floor. Bending is also experienced by pipelines during installation by horizontal directional drilling. HDD procedures are increasingly being utilized around the world not only for crossings of rivers and other obstacles but also for shore approach of offshore pipelines. During installation the pipeline experience a combination of tensile, bending, and compressive stresses. The magnitude of these stresses is a function of the approach angle, bending radius, pipe diameter, length of the borehole, and the soil properties at the site. The objective of this paper is to present an overview of some aspects related to bending of the product pipe during HDD operations, which is closely related to the borehole path as the pipeline conforms to the curvature of the hole. An overview of the aspects related to tensile forces is also presented. The combined effect of bending and tensile forces during the pullback operation is discussed. (author)

Comparison of oxide- and metal-core behavior during CRBRP [Clinch River Breeder Reactor Plant] station blackout

International Nuclear Information System (INIS)

Polkinghorne, S.T.; Atkinson, S.A.

1986-01-01

A resurrected concept that could significantly improve the inherently safe response of Liquid-Metal cooled Reactors (LMRs) during severe undercooling transients is the use of metallic fuel. Analytical studies have been reported on for the transient behavior of metal-fuel cores in innovative, inherently safe LMR designs. This paper reports on an analysis done, instead, for the Clinch River Breeder Reactor Plant (CRBRP) design with the only innovative change being the incorporation of a metal-fuel core. The SSC-L code was used to simulate a protected station blackout accident in the CRBRP with a 943 MWt Integral Fast Reactor (IFR) metal-fuel core. The results, compared with those for the oxide-fueled CRBRP, show that the margin to boiling is greater for the IFR core. However, the cooldown transient is more severe due to the faster thermal response time of metallic fuel. Some additional calculations to assess possible LMR design improvements (reduced primary system pressure losses, extended flow coastdown) are also discussed. 8 refs., 13 figs., 2 tabs

Seismic response analyses for reactor facilities at Savannah River

International Nuclear Information System (INIS)

Miller, C.A.; Costantino, C.J.; Xu, J.

1991-01-01

The reactor facilities at the Savannah River Plant (SRP) were designed during the 1950's. The original seismic criteria defining the input ground motion was 0.1 G with UBC [uniform building code] provisions used to evaluate structural seismic loads. Later ground motion criteria have defined the free field seismic motion with a 0.2 G ZPA [free field acceleration] and various spectral shapes. The spectral shapes have included the Housner spectra, a site specific spectra, and the US NRC [Nuclear Regulatory Commission] Reg. Guide 1.60 shape. The development of these free field seismic criteria are discussed in the paper. The more recent seismic analyses have been of the following type: fixed base response spectra, frequency independent lumped parameter soil/structure interaction (SSI), frequency dependent lumped parameter SSI, and current state of the art analyses using computer codes such as SASSI. The results from these computations consist of structural loads and floor response spectra (used for piping and equipment qualification). These results are compared in the paper and the methods used to validate the results are discussed. 14 refs., 11 figs

The influence of Savannah River discharge and changing SRS cooling water requirements on the potential entrainment of ichthyoplankton at the SRS Savannah River intakes

International Nuclear Information System (INIS)

Paller, M.H.

1992-08-01

Entrainment (i.e., withdrawal of fish larvae and eggs in cooling water) at the SRS Savannah River intakes is greatest when periods of high river water usage coincide with low river dischargeduring the spawning season. American shad and striped bass are the two species of greatest concern because of their recreational and/or commercial importance and because they produce drifting eggs and larvae vulnerable to entrainment. In the mid-reaches of the Savannah River, American shad and striped bass spawn primarily during April and May. An analysis of Savannah River discharge during April and May 1973--1989 indicated the potential for entrainment of 4--18% of the American shad and striped bass larvae and eggs that drifted past the SRS. This analysis assumed the concurrent operation of L-, K-, and P-Reactors. Additional scenarios investigated were: (1) shutting down L- and P-Reactors, and operating K-Reactor with a recycle cooling tower; and (2) shutting down L- and P-Reactors, eliminating minimum flows to Steel Creek, and operating K-Reactor with a recycle cooling tower. The former scenario reduced potential entrainment to 0.7--3.3%, and the latter scenario reduced potential entrainment to 0.20.8%. Thus, the currently favored scenario of operating K-Reactor with a cooling tower and not operating L- and P-Reactors represents a significant lessening of the impact of SRS operations

Reactor BR2

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2000-07-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

Reactor BR2

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

Ichthyoplankton entrainment study at the SRS Savannah River water intakes for Westinghouse Savannah River Company

International Nuclear Information System (INIS)

Paller, M.

1992-01-01

Cooling water for L and K Reactors and makeup water for Par Pond is pumped from the Savannah River at the 1G, 3G, and 5G pump houses. Ichthyoplankton (drifting fish larvae and eggs) from the river are entrained into the reactor cooling systems with the river water and passed through the reactor's heat exchangers where temperatures may reach 70 degrees C during full power operation. Ichthyoplankton mortality under such conditions is assumed to be 100 percent. The number of ichthyoplankton entrained into the cooling system depends on a variety of variables, including time of year, density and distribution of ichthyoplankton in the river, discharge levels in the river, and the volume of water withdrawn by the pumps. Entrainment at the 1 G pump house, which is immediately downstream from the confluence of Upper Three Runs Creek and the Savannah River, is also influenced by discharge rates and ichthyoplankton densities in Upper Three Runs Creek. Because of the anticipated restart of several SRS reactors and the growing concern surrounding striped bass and American shad stocks in the Savannah River, the Department of Energy requested that the Environmental Sciences Section (ESS) of the Savannah River Laboratory sample ichthyoplankton at the SRS Savannah River intakes. Dams ampersand Moore, Inc., under a contract with Westinghouse Savannah River Company performed the sampling and data analysis for the ESS

Geomorphological evidences of Quaternary tectonic activities in the Santa Cruz river valley, Patagonia, Argentina

International Nuclear Information System (INIS)

Massabie, A.; Sanguinetti, A.; Nestiero, O.

2007-01-01

From Argentin lake, at west on Andean hills, to Puerto Santa Cruz on Atlantic coast, Santa Cruz river cross eastward Santa Cruz province over 250 km in Patagonia at southern Argentina. Present bed of the river has a meandering outline with first order meanders of great ratio bends and second order meanders of minor ratio bends. Principal wanderings are 45 to 55 km spaced from near Estancia La Julia or Rio Bote at west to Comandante Luis Piedrabuena at east. On river's bed middle sector these great curvatures are located at Estancia Condor Cliff and Estancia Rincon Grande. Regional and partial detailed studies allow to recognize structural control on river's bed sketch and valley s geomorphology that relates first order bends with reactivated principal faults. These faults fit well with parallel system of northwest strike of Austral Basin.On geological, geomorphologic and structural evidences recognized in Santa Cruz river, quaternary tectonic activity, related to Andean movements in southern Patagonian foreland, is postulated. (author)

Internal fluid flow management analysis for Clinch River Breeder Reactor Plant sodium pumps

International Nuclear Information System (INIS)

Cho, S.M.; Zury, H.L.; Cook, M.E.; Fair, C.E.

1978-12-01

The Clinch River Breeder Reactor Plant (CRBRP) sodium pumps are currently being designed and the prototype unit is being fabricated. In the design of these large-scale pumps for elevated temperature Liquid Metal Fast Breeder Reactor (LMFBR) service, one major design consideration is the response of the critical parts to severe thermal transients. A detailed internal fluid flow distribution analysis has been performed using a computer code HAFMAT, which solves a network of fluid flow paths. The results of the analytical approach are then compared to the test data obtained on a half-scale pump model which was tested in water. The details are presented of pump internal hydraulic analysis, and test and evaluation of the half-scale model test results

Present day design challenges exemplified by the Clinch River Breeder Reactor Plant

International Nuclear Information System (INIS)

Dickson, P.W. Jr.; Anderson, C.A. Jr.

1976-01-01

The present day design challenges faced by the Clinch River Breeder Reactor Plant engineer result from two causes. The first cause is aspiration to achieve a design that will operate at conditions which are desirable for future LMFBRs in order for them to achieve low power costs and good breeding. The second cause is the licensing impact. Although licensing the CRBRP won't eliminate future licensing effort, many licensing questions will have been resolved and precedents set for the future LMFBR industry

Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

International Nuclear Information System (INIS)

Pope, R.B.; Diggs, J.M.

1982-04-01

Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented

Description of the two-loop RELAP5 model of the L-Reactor at the Savannah River Site

International Nuclear Information System (INIS)

Cozzuol, J.M.; Davis, C.B.

1989-12-01

A two-loop RELAP5 input model of the L-Reactor at the Savannah River Site (SRS) was developed to support thermal-hydraulic analysis of SRS reactors. The model was developed to economically evaluate potential design changes. The primary simplifications in the model were in the number of loops and the detail in the moderator tank. The six loops in the reactor were modeled with two loops, one representing a single loop and the other representing five combined loops. The model has undergone a quality assurance review. This report describes the two-loop model, its limitations, and quality assurance. 29 refs., 18 figs., 10 tabs

Record of Decision; Continued operation of K, L, and P Reactors, Savannah River Site, Aiken, South Carolina

International Nuclear Information System (INIS)

1991-01-01

The US Department of Energy (DOE) has considered the environmental impacts, benefits and costs, and institutional and programmatic needs associated with continued operation of the Savannah River Site (SRS) reactors, and has decided that it will continue to operate K and L Reactors at SRS, and will terminate operation of P Reactor in the immediate future and maintain it in cold standby. For P Reactor, this will involve the reactor's defueling; storage of its heavy water moderator in tanks in the reactor building; shutdown of reactor equipment and systems in a protected condition to prevent deterioration; and maintenance of the reactor in a defueled, protected status by a skeleton staff, which would permit any future decision to refuel and restart. Currently committed and planned upgrade activities will be discontinued for P Reactor. DOE will proceed with the safety upgrades and management system improvements currently scheduled for K Reactor in its program to satisfy the criteria of the Safety Evaluation Report (SER), and will conduct an Operational Readiness Review (ORR). The satisfaction of the SER criteria and completion of the ORR will demonstrate that the safety and health criteria for the resumption of production have been met. Reactor restart is expected to be in the third quarter of 1991 for K Reactor

Tilted bending magnet for SPS target area TCC2

CERN Multimedia

CERN PhotoLab

1976-01-01

A slow-extracted proton beam from the SPS goes to the underground target zone TCC2. The part of the primary beam which traverses target T4 is recuperated and transported over some 800 m, for further use in the North Area High Intensity facility (NAHIF). The curved and sloped trajectory required 4 of the bending magnets to be tilted. Here we see one of them being attended by Gilbert Françon in hall 867, ready for installation in TCC2.

Nuclear reactor in deep water

International Nuclear Information System (INIS)

Anon.

1980-01-01

Events during October 1980, when the Indian Point 2 nuclear reactor was flooded by almost 500 000 litres of water from the Hudson river, are traced and the jumble of human errors and equipment failures chronicled. Possible damage which could result from the reactor getting wet and from thermal shock are considered. (U.K.)

Decommissioning an Active Historical Reactor Facility at the Savannah River Site - 13453

Energy Technology Data Exchange (ETDEWEB)

Bergren, Christopher L.; Long, J. Tony; Blankenship, John K. [Savannah River Nuclear Solutions, LLC, Bldg. 730-4B, Aiken, SC 29808 (United States); Adams, Karen M. [United States Department of Energy, Bldg. 730-B, Aiken, SC 29808 (United States)

2013-07-01

The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, where Management and Operations are performed by Savannah River Nuclear Solutions (SRNS). In 2004, DOE recognized SRS as structure within the Cold War Historic District of national, state and local significance composed of the first generation of facilities constructed and operated from 1950 through 1989 to produce plutonium and tritium for our nation's defense. DOE agreed to manage the SRS 105-C Reactor Facility as a potentially historic property due to its significance in supporting the U.S. Cold War Mission and for potential for future interpretation. This reactor has five primary areas within it, including a Disassembly Basin (DB) that received irradiated materials from the reactor, cooled them and prepared the components for loading and transport to a Separation Canyon for processing. The 6,317 square meter area was divided into numerous work/storage areas. The walls between the individual basin compartments have narrow vertical openings called 'slots' that permit the transfer of material from one section to another. Data indicated there was over 830 curies of radioactivity associated with the basin sediments and approximately 9.1 M liters of contaminated water, not including a large quantity of activated reactor equipment, scrap metal, and debris on the basin floor. The need for an action was identified in 2010 to reduce risks to personnel in the facility and to eliminate the possible release of contaminants into the environment. The release of DB water could potentially migrate to the aquifer and contaminate groundwater. DOE, its regulators [U. S. Environmental Protection Agency (USEPA)-Region 4 and the South Carolina Department of Health and Environmental Control (SCDHEC)] and the SC Historical Preservation Office (SHPO) agreed/concurred to perform a non

Decommissioning an Active Historical Reactor Facility at the Savannah River Site - 13453

International Nuclear Information System (INIS)

Bergren, Christopher L.; Long, J. Tony; Blankenship, John K.; Adams, Karen M.

2013-01-01

The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, where Management and Operations are performed by Savannah River Nuclear Solutions (SRNS). In 2004, DOE recognized SRS as structure within the Cold War Historic District of national, state and local significance composed of the first generation of facilities constructed and operated from 1950 through 1989 to produce plutonium and tritium for our nation's defense. DOE agreed to manage the SRS 105-C Reactor Facility as a potentially historic property due to its significance in supporting the U.S. Cold War Mission and for potential for future interpretation. This reactor has five primary areas within it, including a Disassembly Basin (DB) that received irradiated materials from the reactor, cooled them and prepared the components for loading and transport to a Separation Canyon for processing. The 6,317 square meter area was divided into numerous work/storage areas. The walls between the individual basin compartments have narrow vertical openings called 'slots' that permit the transfer of material from one section to another. Data indicated there was over 830 curies of radioactivity associated with the basin sediments and approximately 9.1 M liters of contaminated water, not including a large quantity of activated reactor equipment, scrap metal, and debris on the basin floor. The need for an action was identified in 2010 to reduce risks to personnel in the facility and to eliminate the possible release of contaminants into the environment. The release of DB water could potentially migrate to the aquifer and contaminate groundwater. DOE, its regulators [U. S. Environmental Protection Agency (USEPA)-Region 4 and the South Carolina Department of Health and Environmental Control (SCDHEC)] and the SC Historical Preservation Office (SHPO) agreed/concurred to perform a non-time critical removal

Radiometric analyses of floodplain sediments at the Savannah River Plant

International Nuclear Information System (INIS)

Lower, M.W.

1987-09-01

A Comprehensive Cooling Water Study to assess the effects of reactor cooling water discharges and related reactor area liquid releases to onsite streams and the nearby Savannah River has been completed at the US Department of Energy's Savannah River Plant (SRP). Extensive radiometric analyses of man-made and naturally occurring gamma-emitting radionuclides were measured in floodplain sediment cores extracted from onsite surface streams at SRP and from the Savannah River. Gamma spectrometric analyses indicate that reactor operations contribute to floodplain radioactivity levels slightly higher than levels associated with global fallout. In locations historically unaffected by radioactive releases from SRP operations, Cs-137 concentrations were found at background and fallout levels of about 1 pCi/g. In onsite streams that provided a receptor for liquid radioactive releases from production reactor areas, volume-weighted Cs-137 concentrations ranged by core from background levels to 55 pCi/g. Savannah River sediments contained background and atmospheric fallout levels of Cs-137 only. 2 refs., 5 figs

Savannah River Site production reactor safety analysis report

International Nuclear Information System (INIS)

1996-01-01

The process water system (PWS) is designed to remove heat produced in the reactor from the fission process, gamma radiation absorption, and fission product decay. Heat removal is accomplished by circulating heavy water through the reactor. Cooling is provided for fuel assemblies, target assemblies, control rods, bulk moderator, deflector plate, reactor tank, and reactor structural components. Approximately 90% of the heat load is generated in the fuel and target assemblies, 5% in the moderator, and 5% in the shielding. In addition to serving as the-heat transfer medium, the process water moderates neutrons produced by fission in the fuel. D 2 O is used in this application because of its favorable moderating and neutron capture properties, which result in high neutron efficiency and reactor productivity. The PWS piping and components also provide a high-integrity leak barrier against loss of moderator and the radioactive fission and corrosion products. Components of the PWS are located in the reactor building between the -40-foot elevation and the 0-foot elevation. Specific locations include the process room, heat exchanger bay, motor rooms, and pump rooms. The system diagram is shown in Figure 5.1-2. PWS design data are presented in Table 5.1-1. The PWS consists of six parallel heat transfer loops. In each loop, approximately 25,000 gpm of D 2 O is circulated from one of six outlet nozzles in the bottom of the reactor tank through a motor-operated valve (MOV) to the suction side of the process water pump. Each pump is driven by an AC motor and a DC motor through a gear reducer unit. A 3-ton flywheel on the drive shaft of the AC motor provides gradual flow coastdown when power is lost. During reactor operation, the DC motors are operated continuously from the diesel generator sets as backup to the AC motors. Following shutdown, the DC motors are operated to provide adequate circulation and core cooling

Dissipation of the reactor heat at the Savannah River Plant

International Nuclear Information System (INIS)

Neill, J.S.; Babcock, D.F.

1971-10-01

The effluent cooling water from the heat exchangers of the Savannah River nuclear reactors is cooled by natural processes as it flows through the stream beds, canals, ponds, and swamps on the plant site. The Langhaar equation, which gives the rate of heat removal from the water surface as a function of the surface temperature, air temperature, relative humidity, and wind speed, is applied satisfactorily to calculate the cooling that occurs at all temperature levels and for all modes of water flow. The application of this equation requires an accounting of effects such as solar heating, shading, mixing, staging, stratification, underflow, rainfall, the imposed heat load, and the rate of change in heat content of the body of water

In-plane and out-of-plane bending tests on carbon steel pipe bends

International Nuclear Information System (INIS)

Brouard, D.; Tremblais, A.; Vrillon, B.

1979-01-01

The objectives of these tests were to obtain experimental results on bends behaviour in elastic and plastic regime by in plane and out of plane bending. Results were used to improve the computer model, for large distorsion of bends, to be used in a simplified beam type computer code for piping calculations. Tests were made on type ANSI B 169 DN 5 bends in ASTM A 106 Grade B carbon steel. These tests made it possible to measure, for identical bends, in elastic regime, the flexibility factors and, in plastic regime, the total evolution in opening, in closing and out of plane. Flexibility factors of 180 0 bend without flanges are approximately the same in opening and in closing. The end effect due to flanges is not very significant, but it is important for 90 0 bends. In plastic regime, collapse loads or collapse moments of bends depends also of both the end effects and the angle bend. The end effects and the angle bend are more sensitive in opening than in closing. The interest of these tests is to procure some precise evolution curves of identical bends well characterized in geometry and metal strength, deflected in large distorsions. (orig./HP)

Evolution of a meander in a constricted reach of a dryland alluvial channel: Little Colorado River, Arizona

Science.gov (United States)

Block, D.

2013-12-01

Lateral migration of river meander systems is complex, particularly in drylands where fluvial processes are discontinuous. Analysis of aerial photography and GPS tracking of cutbank erosion can further empirical knowledge of meander development. Moreover, discharge records link landscape response to hydroclimatic variability. In the semiarid Little Colorado River valley, extreme erosive episodes typically result from snowmelt flow, or lately, rain-on-snow events. The 90-km reach of the Little Colorado River (LCR), from Winslow to Leupp, Arizona, meanders within a 5-km-wide valley. Near Winslow, however, the LCR is disconnected from its floodplain by a 12-km-long levee. The levee restricts the floodplain to only 450 m wide in one location. In this severely constricted river stretch, a flood event in January 2008 relocated a meander bend. Bend development followed a common sequence of migration phases long noted in the literature, but at a very rapid pace. During the flood event one meander limb migrated ~200 m, following the general northwesterly flow direction of the river. Movement vectors of meander inflection points, apex, and apical line characterize changes in bend morphology. Before the 2008 flood event the apical line of the meander bend had azimuth 50°; after the 2008 flood event the apical line of the meander bend had azimuth 345°. Since that event, the meander bend has migrated an additional ~200 m through a combination of translation, extension, and rotation. The data provide information on geomorphic response to bimodal precipitation patterns in a human-perturbed channel reach.

Review of reports associated with systems of the K, P and L reactors at the Savannah River Site

International Nuclear Information System (INIS)

Cowgill, M.G.

1992-02-01

Six reports associated with the structural integrity of several systems of the Savannah River Site reactors are reviewed. The focus is on the materials-related aspects of the reports and no attempt is made to address the stress analysis-related issues

Effect of interlayer bonding strength and bending stiffness on 2-dimensional materials’ frictional properties at atomic-scale steps

International Nuclear Information System (INIS)

Lang, Haojie; Peng, Yitian; Zeng, Xingzhong

2017-01-01

Highlights: • Bending of uncovered step edge of 2-dimensional materials could be a common phenomenon during friction processes. • 2-dimensional materials with large interlayer bonding strength possess good frictional properties at step. • Increased bending stiffness of step edge could be the major reason that lateral force increased with step height. - Abstract: Atomic-scale steps generally presented in 2-dimensional materials have important influence on the overall nanotribological properties of surface. Frictional properties at atomic-scale steps of two types of 2-dimensional materials are studied using calibrated atomic force microscopy (AFM) tip sliding against the steps. The lateral force at uncovered step is larger than covered step due to the bending of step edge. The lateral force at monolayer uncovered step edge of h-BN is lower than graphene because h-BN possesses higher interlayer bonding strength than graphene and the bending of h-BN step edge is suppressed to some extent. The high uncovered step exhibits much larger lateral force than low uncovered step, which could be mainly induced by increased bending stiffness of step edge rather than increased step height. The results revealed that interlayer bonding strength and bending stiffness have great influence on the lateral force at atomic-scale steps. The studies can provide a further understanding of frictional properties at atomic scale steps and could be helpful for the applications of 2-dimensional materials as lubricant coating.

Effect of interlayer bonding strength and bending stiffness on 2-dimensional materials’ frictional properties at atomic-scale steps

Energy Technology Data Exchange (ETDEWEB)

Lang, Haojie; Peng, Yitian, E-mail: yitianpeng@dhu.edu.cn; Zeng, Xingzhong

2017-07-31

Highlights: • Bending of uncovered step edge of 2-dimensional materials could be a common phenomenon during friction processes. • 2-dimensional materials with large interlayer bonding strength possess good frictional properties at step. • Increased bending stiffness of step edge could be the major reason that lateral force increased with step height. - Abstract: Atomic-scale steps generally presented in 2-dimensional materials have important influence on the overall nanotribological properties of surface. Frictional properties at atomic-scale steps of two types of 2-dimensional materials are studied using calibrated atomic force microscopy (AFM) tip sliding against the steps. The lateral force at uncovered step is larger than covered step due to the bending of step edge. The lateral force at monolayer uncovered step edge of h-BN is lower than graphene because h-BN possesses higher interlayer bonding strength than graphene and the bending of h-BN step edge is suppressed to some extent. The high uncovered step exhibits much larger lateral force than low uncovered step, which could be mainly induced by increased bending stiffness of step edge rather than increased step height. The results revealed that interlayer bonding strength and bending stiffness have great influence on the lateral force at atomic-scale steps. The studies can provide a further understanding of frictional properties at atomic scale steps and could be helpful for the applications of 2-dimensional materials as lubricant coating.

Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

International Nuclear Information System (INIS)

Owen, M.B.

1996-08-01

This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR

A Numerical Model for Flow and Sediment Transport in Alluvial-River Bends.

Science.gov (United States)

1983-12-01

into the bend divided by the computed total sediment discharge across the section. This insures that sediment continuity is preserved along the...6 * * * 0 * C YORF DEFINITION’S , C PI : PI !!! BOSTON CREME, APPLE, PUMPKIN , LTC* "" C G = GRAVITATIONAL CONSTANT * r AAA = COEFFICI92T IN

Keeping research reactors relevant: A pro-active approach for SLOWPOKE-2

International Nuclear Information System (INIS)

Cosby, L.R.; Bennett, L.G.I.; Nielsen, K.; Weir, R.

2010-01-01

The SLOWPOKE is a small, inherently safe, pool-type research reactor that was engineered and marketed by Atomic Energy of Canada Limited (AECL) in the 1970s and 80s. The original reactor, SLOWPOKE-1, was moved from Chalk River to the University of Toronto in 1970 and was operated until upgraded to the SLOWPOKE-2 reactor in 1973. In all, eight reactors in the two versions were produced and five are still in operation today, three having been decommissioned. All of the remaining reactors are designated as SLOWPOKE-2 reactors. These research reactors are prone to two major issues: aging components and lack of relevance to a younger audience. In order to combat these problems, one SLOWPOKE -2 facility has embraced a strategy that involves modernizing their reactor in order to keep the reactor up to date and relevant. In 2001, this facility replaced its aging analogue reactor control system with a digital control system. The system was successfully commissioned and has provided a renewed platform for student learning and research. The digital control system provides a better interface and allows flexibility in data storage and retrieval that was never possible with the analogue control system. This facility has started work on another upgrade to the digital control and instrumentation system that will be installed in 2010. The upgrade includes new computer hardware, updated software and a web-based simulation and training system that will allow licensed operators, students and researchers to use an online simulation tool for training, education and research. The tool consists of: 1) A dynamic simulation for reactor kinetics (e.g., core flux, power, core temperatures, etc). This tool is useful for operator training and student education; 2) Dynamic mapping of the reactor and pool container gamma and neutron fluxes as well as the vertical neutron beam tube flux. This research planning tool is used for various researchers who wish to do irradiations (e.g., neutron

Analysis of ductile-brittle transition shifts for standard and miniature bending specimens of irradiated steel

International Nuclear Information System (INIS)

Korshunov, M.E.; Korolev, Yu.N.; Krasikov, E.A.; Gabuev, N.N.; Tykhmeev, D.Yu.

1996-01-01

A study is made to reveal if there is a correlation between shifts in temperature curves obtained when testing thin plates and standard specimens on impact bending and fracture toughness. The tests were carried out using steel 25Kh3NM specimens irradiated by 6 x 10 19 cm -2 neutron fluence. A conclusion is made about the possibility to evaluate the degree of radiation-induced embrittlement of reactor steels on the basis of thin plate testing under quasistatic loads [ru

Extensions to SCDAP/RELAP5/MOD2 debris analysis models for the severe accident analysis of Savannah River Site (SRS) reactors preliminary design report

International Nuclear Information System (INIS)

Siefken, L.J.; Moore, R.L.

1989-06-01

Proposed extensions to the debris analysis model in the SCDAP/RELAP5 code to perform severe accident analyses of Savannah River Plant reactors are described. Designs are presented for the following areas of development: (a) calculating convective and radiative heat transfer at the surfaces of a debris region; (b) calculating heatup of a structure and supported debris that interfaces with several fluid control volumes; (c) modeling the addition of transported material to the surfaces of any structure represented by the debris analysis model; (d) calculating the two-dimensional heatup of an arbitrary number of structures in the reactor system; (e) modeling the effect of natural convection of liquefied material on heat transfer in a debris bed; and (f) modeling fission product release and aerosol generation in a debris bed. 11 refs., 12 figs., 7 tabs

Computation of the flow in shallow river bends

NARCIS (Netherlands)

Kalkwijk, J.P.T.; De Vriend, H.J.

1980-01-01

The mathematical model presented describes the flow in rivers of which: i the depth is small compared with the width, ii the width is small compared with the radius of curvature, iii the horizontal length scale of the bottom variations is of the order of magnitude of the width. Within these limits,

A transparent bending-insensitive pressure sensor

Science.gov (United States)

Lee, Sungwon; Reuveny, Amir; Reeder, Jonathan; Lee, Sunghoon; Jin, Hanbit; Liu, Qihan; Yokota, Tomoyuki; Sekitani, Tsuyoshi; Isoyama, Takashi; Abe, Yusuke; Suo, Zhigang; Someya, Takao

2016-05-01

Measuring small normal pressures is essential to accurately evaluate external stimuli in curvilinear and dynamic surfaces such as natural tissues. Usually, sensitive and spatially accurate pressure sensors are achieved through conformal contact with the surface; however, this also makes them sensitive to mechanical deformation (bending). Indeed, when a soft object is pressed by another soft object, the normal pressure cannot be measured independently from the mechanical stress. Here, we show a pressure sensor that measures only the normal pressure, even under extreme bending conditions. To reduce the bending sensitivity, we use composite nanofibres of carbon nanotubes and graphene. Our simulations show that these fibres change their relative alignment to accommodate bending deformation, thus reducing the strain in individual fibres. Pressure sensitivity is maintained down to a bending radius of 80 μm. To test the suitability of our sensor for soft robotics and medical applications, we fabricated an integrated sensor matrix that is only 2 μm thick. We show real-time (response time of ∼20 ms), large-area, normal pressure monitoring under different, complex bending conditions.

SiC-CMC-Zircaloy-4 Nuclear Fuel Cladding Performance during 4-Point Tubular Bend Testing

Energy Technology Data Exchange (ETDEWEB)

IJ van Rooyen; WR Lloyd; TL Trowbridge; SR Novascone; KM Wendt; SM Bragg-Sitton

2013-09-01

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for “accident tolerant” nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designs being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between loading points

Processes influencing cooling of reactor effluents

International Nuclear Information System (INIS)

Magoulas, V.E.; Murphy, C.E. Jr.

1982-01-01

Discharge of heated reactor cooling water from SRP reactors to the Savannah River is through sections of stream channels into the Savannah River Swamp and from the swamp into the river. Significant cooling of the reactor effluents takes place in both the streams and swamp. The majority of the cooling is through processes taking place at the surface of the water. The major means of heat dissipation are convective transfer of heat to the air, latent heat transfer through evaporation and radiative transfer of infrared radiation. A model was developed which incorporates the effects of these processes on stream and swamp cooling of reactor effluents. The model was used to simulate the effect of modifications in the stream environment on the temperature of water flowing into the river. Environmental effects simulated were the effect of changing radiant heat load, the effect of changes in tree canopy density in the swamp, the effect of total removal of trees from the swamp, and the effect of diverting the heated water from L reactor from Steel Creek to Pen Branch. 6 references, 7 figures

Photoelastic investigation of the stresses in mitred bent-cylinders under bending, 2

International Nuclear Information System (INIS)

Sawa, Yoshiaki

1984-01-01

The states of stress distribution in mitred bend subjected to inplane uniform bending moment have been studied systematically by means of photoelastic stress freezing method. The relations between the stress concentration factor of fiber stress σsub(l) and of hoop stress σsub(theta) near the bent part and the angle of mitred bend are thoroughly investigated. The effects of fillet radius of the bent-part and wall thickness on the stress concentration factors are also discussed. (author)

Experimental Study of the Effect of W-weir on Reduction of Scour Depth at 90 Degree Sharp Bend

Directory of Open Access Journals (Sweden)

Vida Atashi

2017-02-01

Full Text Available Introduction: Flow patterns within the river bend is three dimensional. Occurrence of secondary flow due to centrifugal force and formation of helicoidally vortex in river bend usually causes the outer bank of river erodes whilst the sediment are deposited in inner bend which appears in the form of point bars. To reduce the river bank scour, many techniques have been developed which may be classified as covering technique and modified flow patterns methods. The W-weir is among such structures. In the present paper, by measuring three components of flow velocity with and without presence of W-weir, variation of flow patterns and shear stress distribution in a 90-degree sharp bend have been investigated. The main purpose of this study is to see the installation of different locations of W-weir in the bend on reduction of outer bank scour. In the present paper, by measuring three components of flow velocity with and without presence of W-weir, variation of flow patterns and shear stress distribution in a 90-degree sharp bend have been investigated. The analyses of data showed more uniform flow upstream of the weir and also revealed that the effect of transverse and centrifugal forces are modified in such a way that the secondary flow is diminished. The results showed that for 30, 60 and 90-degree bends maximum erosion depth in the vicinity of the outer bank with Froude number of 0.206 in comparison with 0.137 has increased up to 84, 90 and 118 % respectively. In both Froude numbers, installation of W-Weir in 30 degree has the most reduction in bed in comparison with 60 and 90 degree. Materials and Methods: To reach the goal of this study a physical model of 90 degree sharp bend was constructed in the hydraulic lab of Shahid Chamran university of Ahvaz. The ratio of R(radius/b(flume width was less than 2 which shows a sharp bend. The W-weir was built with 1mm galvanized steel. Flume bed was covered with sediment of D50=1.5mm. The W-weir was

Geotechnical investigation for seismic issues for K-reactor area at Savannah River Site

International Nuclear Information System (INIS)

Castro, G.; Reeves, C.Q.

1991-01-01

A geotechnical investigation has been completed at Savannah River Site to characterize the foundation conditions in K-Reactor Area and confirm soil design properties for use in seismic qualification of structures. The scope of field work included ten soil borings to a 200-foot depth with split-spoon and undisturbed sampling. Additionally, 42 cone penetrometer tests were performed with seismic down-hole measurements. Three cross-hole shear wave velocity tests were also completed to confirm the assumed dynamic properties which had been used in preliminary seismic analysis

Adjustable Tooling for Bending Brake

Science.gov (United States)

Ellis, J. M.

1986-01-01

Deep metal boxes and other parts easily fabricated. Adjustable tooling jig for bending brake accommodates spacing blocks and either standard male press-brake die or bar die. Holds spacer blocks, press-brake die, bar window die, or combination of three. Typical bending operations include bending of cut metal sheet into box and bending of metal strip into bracket with multiple inward 90 degree bends. By increasing free space available for bending sheet-metal parts jig makes it easier to fabricate such items as deep metal boxes or brackets with right-angle bends.

A Numerical Study of the Spring-Back Phenomenon in Bending with a Rebar Bending Machine

Directory of Open Access Journals (Sweden)

Chang Hwan Choi

2014-10-01

Full Text Available Recently, the rebar bending methodology started to change from field processing to utilizing rebar bending machines at plant sites prior to transport to the construction locations. Computerized control of rebar plant bending machines provides more accurate and faster bending of rebars than the low quality inefficient field processing alternative. The bending process involves plastic deformation of rebars, where bending stress beyond the yield point of the material is applied. When the bending stress is removed, spring back is caused by the elastic restoring stress. Therefore, an accurate numerical analysis of the spring-back process is required to reduce the bending process errors. The most sensitive factors affecting the spring-back process are the bending radius, the bending angle, the diameter of the rebar, the friction coefficient, and the yielding strength of material. In this paper, we suggest a numerical modeling method using these factors. The finite element modeling of the dynamic mechanical behavior of the material during bending is performed using a commercial dynamic analysis program “DAFUL.” We use the least squares approach to derive the spring-back deflection as a function of the rebar bending parameters.

"Wandering in the Desert": The Clinch River Breeder Reactor Debate in the U.S. Congress, 1972-1983.

Science.gov (United States)

Camp, Michael

2018-01-01

The experimental Clinch River breeder reactor, approved by the U.S. Congress in 1970 for construction in East Tennessee, would have used plutonium instead of uranium. The project drew the ire of environmentalists who insisted that plutonium was too dangerous for commercial use, along with opponents of nuclear proliferation. Tennessee's representatives in Congress, however, desired the jobs that the project would create, and formed legislative coalitions to ensure continued appropriations for the project. Funding lasted until 1983, when fiscal conservatives, concerned about ballooning cost projections, joined with environmentalists to defund the breeder. Interpretations of U.S. nuclear policy in the 1980s have often revolved around the Three Mile Island meltdown's aftermath, but Clinch River was not affected by the incident. Instead, the Clinch River controversy revolved around other unrelated issues. The Clinch River story therefore offers a corrective to accounts that privilege national public opinion at the expense of other variables.

Analytic description of the frictionally engaged in-plane bending process incremental swivel bending (ISB)

Science.gov (United States)

Frohn, Peter; Engel, Bernd; Groth, Sebastian

2018-05-01

Kinematic forming processes shape geometries by the process parameters to achieve a more universal process utilizations regarding geometric configurations. The kinematic forming process Incremental Swivel Bending (ISB) bends sheet metal strips or profiles in plane. The sequence for bending an arc increment is composed of the steps clamping, bending, force release and feed. The bending moment is frictionally engaged by two clamping units in a laterally adjustable bending pivot. A minimum clamping force hindering the material from slipping through the clamping units is a crucial criterion to achieve a well-defined incremental arc. Therefore, an analytic description of a singular bent increment is developed in this paper. The bending moment is calculated by the uniaxial stress distribution over the profiles' width depending on the bending pivot's position. By a Coulomb' based friction model, necessary clamping force is described in dependence of friction, offset, dimensions of the clamping tools and strip thickness as well as material parameters. Boundaries for the uniaxial stress calculation are given in dependence of friction, tools' dimensions and strip thickness. The results indicate that changing the bending pivot to an eccentric position significantly affects the process' bending moment and, hence, clamping force, which is given in dependence of yield stress and hardening exponent. FE simulations validate the model with satisfactory accordance.

Closed-form plastic collapse loads of pipe bends under combined pressure and in-plane bending

International Nuclear Information System (INIS)

Oh, Chang Sik; Kim, Yun Jae

2006-01-01

Based on three-dimensional (3-D) FE limit analyses, this paper provides plastic limit, collapse and instability load solutions for pipe bends under combined pressure and in-plane bending. The plastic limit loads are determined from FE limit analyses based on elastic-perfectly plastic materials using the small geometry change option, and the FE limit analyses using the large geometry change option provide plastic collapse loads (using the twice-elastic-slope method) and instability loads. For the bending mode, both closing bending and opening bending are considered, and a wide range of parameters related to the bend geometry is considered. Based on the FE results, closed-form approximations of plastic limit and collapse load solutions for pipe bends under combined pressure and bending are proposed

Safety features of the MAPLE-X10 reactor design

International Nuclear Information System (INIS)

Lee, A.G.; Bishop, W.E.; Heeds, W.

1990-09-01

The MAPLE-X10 reactor is a D 2 0-reflected, H 2 0-cooled and -moderated pool-type reactor under construction at the Chalk River Nuclear Laboratories. This 10-MW reactor will produce key medical and industrial radio-isotopes such as 99 Mo, 125 I, and 192 Ir. As the prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor since standards for the licensing of new research reactors have not been developed yet by the licensing authority in Canada

Restart of R reactor at SRP

International Nuclear Information System (INIS)

McDonell, W.R.

1983-01-01

Restart of the Savannah River R-Reactor is an alternative to L-Reactor operation for increased production of defense nuclear material. R-Reactor was shut down in 1964 after 11-years operation and has been on standby for 19 years. This report presents a description of R-Reactor operation to serve as a basis for analysis of environmental impacts after restoration to meet current SRP performance standards. R-Reactor operation would differ from L-Reactor operation principally in discharge and recycle of effluent cooling water to Par Pond, rather than direct discharge to the Savannah River by way of Steel Creek. Significant differences in environmental effects could result. A costly renovation program would be required to restore R-Reactor to operability, and the reactor could not contribute to material production before about 1989

Seismic design criteria for the Clinch River Breeder Reactor Plant

International Nuclear Information System (INIS)

Morrone, A.; Bitner, J.L.; Sigal, G.B.

1975-01-01

The general criteria for seismic resistant design for structures, systems and components of the Clinch River Breeder Reactor Plant (CRBRP) are presented and discussed. Site dependency of the maximum ground accelerations for the Operating Basis Earthquake and the Safe Shutdown Earthquake is described from the viewpoint of historical records and geological and seismological studies for the CRBRP site. The respective ground response spectra are derived by normalization of the latest AEC Regulatory standard shapes to these maximum ground accelerations. Modeling and analytical techniques and requirements are given. In addition, loading conditions and categories, loading combinations, earthquake direction effects and allowable damping values are defined. A discussion of the testing criteria which considers both single and multiple frequency test motions, and basic test procedures for single frequency sine beat testing is presented. (U.S.)

Protected air-cooled condenser for the Clinch River Breeder Reactor Plant

International Nuclear Information System (INIS)

Louison, R.; Boardman, C.E.

1981-01-01

The long term residual heat removal for the Clinch River Breeder Reactor Plant (CRBRP) is accomplished through the use of three protected air-cooled condensers (PACC's) each rated at 15M/sub t/ following a normal or emergency shutdown of the reactor. Steam is condensed by forcing air over the finned and coiled condenser tubes located above the steam drums. The steam flow is by natural convection. It is drawn to the PACC tube bundle for the steam drum by the lower pressure region in the tube bundle created from the condensing action. The concept of the tube bundle employs a unique patented configuration which has been commercially available through CONSECO Inc. of Medfore, Wisconsin. The concept provides semi-parallel flow that minimizes subcooling and reduces steam/condensate flow instabilities that have been observed on other similar heat transfer equipment such as moisture separator reheaters (MSRS). The improved flow stability will reduce temperature cycling and associated mechanical fatigue. The PACC is being designed to operate during and following the design basis earthquake, depressurization from the design basis tornado and is housed in protective building enclosure which is also designed to withstand the above mentioned events

Large Eddy Simulation of Supercritical CO2 Through Bend Pipes

Science.gov (United States)

He, Xiaoliang; Apte, Sourabh; Dogan, Omer

2017-11-01

Supercritical Carbon Dioxide (sCO2) is investigated as working fluid for power generation in thermal solar, fossil energy and nuclear power plants at high pressures. Severe erosion has been observed in the sCO2 test loops, particularly in nozzles, turbine blades and pipe bends. It is hypothesized that complex flow features such as flow separation and property variations may lead to large oscillations in the wall shear stresses and result in material erosion. In this work, large eddy simulations are conducted at different Reynolds numbers (5000, 27,000 and 50,000) to investigate the effect of heat transfer in a 90 degree bend pipe with unit radius of curvature in order to identify the potential causes of the erosion. The simulation is first performed without heat transfer to validate the flow solver against available experimental and computational studies. Mean flow statistics, turbulent kinetic energy, shear stresses and wall force spectra are computed and compared with available experimental data. Formation of counter-rotating vortices, named Dean vortices, are observed. Secondary flow pattern and swirling-switching flow motions are identified and visualized. Effects of heat transfer on these flow phenomena are then investigated by applying a constant heat flux at the wall. DOE Fossil Energy Crosscutting Technology Research Program.

Plastic loads of pipe bends under combined pressure and out-of-plane bending

International Nuclear Information System (INIS)

Lee, Kuk Hee; Kim, Yun Jae; Park, Chi Yong; Lee, Sung Ho; Kim, Tae Ryong

2007-01-01

Based on three-Dimensional (3-D) FE limit analyses, this paper provides plastic limit and TES(Twice- Elastic-Slope) loads for pipe bends under combined pressure and out-of-plane bending. The plastic limit loads are determined from FE limit analyses based on elastic.perfectly-plastic materials using the small geometry change option, and the FE limit analyses using the large geometry change option provide TES plastic loads. A wide range of parameters related to the bend geometry is considered. Based on the FE results, closed-form approximations of plastic limit and TES plastic load solutions for pipe bends under out-of-plane bending are proposed

TERAHERTZ SPECTROSCOPY AND GLOBAL ANALYSIS OF THE BENDING VIBRATIONS OF ACETYLENE 12C2D2

International Nuclear Information System (INIS)

Yu Shanshan; Drouin, Brian J.; Pearson, John C.; Pickett, Herbert M.; Lattanzi, Valerio; Walters, Adam

2009-01-01

Two hundred and fifty-one 12 C 2 D 2 transitions have been measured in the 0.2-1.6 THz region of its ν 5 -ν 4 difference band and 202 of them were observed for the first time. The accuracy of these measurements is estimated to be ranging from 50 kHz to 100 kHz. The 12 C 2 D 2 molecules were generated under room temperature by passing 120-150 mTorr D 2 O vapor through calcium carbide (CaC 2 ) powder. A multistate analysis was carried out for the bending vibrational modes ν 4 and ν 5 of 12 C 2 D 2 , which includes the lines observed in this work and prior microwave, far-infrared and infrared data on the pure bending levels. Significantly improved molecular parameters were obtained for 12 C 2 D 2 by adding the new measurements to the old data set, which had only 10 lines with microwave measurement precision. New frequency and intensity predictions have been made based on the obtained molecular parameters. The more precise measurements and new predictions reported here will support the analyses of astronomical observations by the future high-resolution spectroscopy telescopes such as Herschel, SOFIA, and ALMA, which will work in the terahertz spectral region.

Estimates of plastic loads for pipe bends under combined in-plane and out-of-plane bending moment

International Nuclear Information System (INIS)

Kim, Nak Hyun; Oh, Chang Sik; Kim, Yun Jae

2008-01-01

This paper provides a method to estimate plastic loads (defined by twice-elastic-slope) for pipe bends under combined in-plane and out-of-plane bending moment, based on detailed 3-D FE limit analyses using elastic-perfectly plastic materials. Because closing bending moment is always lower than opening bending moment, the combination of in-plane closing bending and out-of-plane bending moment becomes the most significant case. Due to conservatism of each bending moments, the resultant moment provided by ASME B and PV code is unduly conservative. However, the concept of the resultant moment is still valid. In this paper, FE results show that the accurate solutions of bending moments provide better estimates of plastic loads of pipe bend under combined in-plane bending and out-of-plane bending moment

Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

Energy Technology Data Exchange (ETDEWEB)

Magoulas, V. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-07-28

Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium, and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.

Assessment of potential impact of the Clinch River Breeder Reactor Plant thermal effluent on the Watts Bar Reservoir striped bass population

International Nuclear Information System (INIS)

Heuer, J.H.; McIntosh, D.; Ostrowski, P.; Tomljanovich, D.A.

1983-11-01

This report is an assessment of potential adverse impact to striped bass (Morone saxatilis) in Watts Bar Reservoir caused by thermal effluent from operation of the Clinch River Breeder Reactor Plant (CRBRP). The Clinch River arm of Watts Bar Reservoir is occupied by adult striped bass during the warmest months of the year. Concern was raised that operation of the CRBRP, specifically thermal discharges, could conflict with management of striped bass. In all cases examined the thermal plume becomes nearly imperceptible within a short distance from the discharge pipe (about 30 ft [10 m]) compared to river width (about 630 ft [190 m]). Under worst case conditions any presence of the plume in the main channel (opposite side of the river from the discharge) will be confined to the surface layer of the water. An ample portion of river cross sections containing ambient temperature water for passage or residence of adult striped bass will always be available in the vicinity of this thermal effluent. Although a small portion of river cross section would exceed the thermal tolerance of striped bass, the fish would naturally avoid this area and seek out adjacent cooler water. Therefore, it is concluded the CRBRP thermal effluent will not significantly affect the integrity of the striped bass thermal refuge in the Clinch River arm of Watts Bar Reservoir. At this time there is no need to consider alternative diffuser designs and thermal modeling. 8 references, 3 figures, 2 tables

Safety features of the MAPLE-X10 reactor design

International Nuclear Information System (INIS)

Lee, A.G.; Bishop, W.E.; Heeds, W.

1990-01-01

This paper reports on the MAPLE-X10 reactor D 2 O-reflected, H 2 O-cooled and -moderated pool- type reactor, under construction at the Chalk River Nuclear Laboratories. This 10-MW will produce key medical and industrial radioisotopes such as 99 Mo, 125 I, and 192 Ir. The prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor as standards for the licensing of new research reactors have not been developed by the licensing authority in Canada

Database for the geologic map of the Bend 30- x 60-minute quadrangle, central Oregon

Science.gov (United States)

Koch, Richard D.; Ramsey, David W.; Sherrod, David R.; Taylor, Edward M.; Ferns, Mark L.; Scott, William E.; Conrey, Richard M.; Smith, Gary A.

2010-01-01

The Bend 30- x 60-minute quadrangle has been the locus of volcanism, faulting, and sedimentation for the past 35 million years. It encompasses parts of the Cascade Range and Blue Mountain geomorphic provinces, stretching from snowclad Quaternary stratovolcanoes on the west to bare rocky hills and sparsely forested juniper plains on the east. The Deschutes River and its large tributaries, the Metolius and Crooked Rivers, drain the area. Topographic relief ranges from 3,157 m (10,358 ft) at the top of South Sister to 590 m (1,940 ft) at the floor of the Deschutes and Crooked Rivers where they exit the area at the north-central edge of the map area. The map encompasses a part of rapidly growing Deschutes County. The city of Bend, which has over 70,000 people living in its urban growth boundary, lies at the south-central edge of the map. Redmond, Sisters, and a few smaller villages lie scattered along the major transportation routes of U.S. Highways 97 and 20. This geologic map depicts the geologic setting as a basis for structural and stratigraphic analysis of the Deschutes basin, a major hydrologic discharge area on the east flank of the Cascade Range. The map also provides a framework for studying potentially active faults of the Sisters fault zone, which trends northwest across the map area from Bend to beyond Sisters. This digital release contains all of the information used to produce the geologic map published as U.S. Geological Survey Geologic Investigations Series I-2683 (Sherrod and others, 2004). The main component of this digital release is a geologic map database prepared using ArcInfo GIS. This release also contains files to view or print the geologic map and accompanying descriptive pamphlet from I-2683.

Prediction of Curve Correction Using Alternate Level Pedicle Screw Placement in Patients With Adolescent Idiopathic Scoliosis (AIS) Lenke 1 and 2 Using Supine Side Bending (SB) and Fulcrum Bending (FB) Radiograph.

Science.gov (United States)

Kwan, Mun Keong; Zeyada, Hassan E; Chan, Chris Yin Wei

2015-10-15

Prospective cohort study. To compare side bending (SB) and fulcrum bending (FB) radiographs in patients with adolescent idiopathic scoliosis (AIS) and effect of magnitude and AR curves on curve correctability. The prediction of correction using side bending flexibility (SBF) and fulcrum bending flexibility (FBF) in alternate level pedicle screw (PS) configuration and effect of curve magnitude and AR curves are not well understood. 100 AIS Lenke 1 and 2 were recruited. Curve magnitude was stratified to G1 (41°-60°), G2 (61°-80°), G3 (>80°). The main thoracic (MT) curves were subclassified to AR curves [Miyanji F, Pawelek JB, Van Valin SE, et al. Is the lumbar modifier useful in surgical decision making? Defining two distinct Lenke 1A curve patterns. Spine 2008;33:2545-51]. Preoperatively SBF and FBF were determined whereas postoperative parameters were correction rate (CR), fulcrum bending correction index (FBCI), and side bending correction index (SBCI). Correlation test were carried out between SBF, FBF versus CR for the cohort. There were 38 (G1), 42 (G2), and 20 (G3) patients. 34% were AR curves. SBF for G1, G2, and G3 were 61.3 ± 14.4, 59.2 ± 16.2 and 43.1 ± 13.1% (P = 0.000) whereas FBF for G1, G2, and G3 were 71.1 ± 16.5, 58.3 ± 18.1 and 52.7 ± 17.1% (P = 0.000). The CR was G1 (74.5 ± 11.5%), G2 (69.2 ± 12.7%), and G3 (70.2 ± 8.6%). FBCI was 1.11 ± 0.3 (G1), 1.28 ± 0.4 (G2) and 1.48 ± 0.6 for G3. SBCI was 1.26 ± 0.2 (G1), 1.50 ± 0.5 (G2), and 1.72 ± 0.4 for G3. There was strong correlation for SBF and FBF versus CR for G1 and G2. For G3, a very strong correlation was established between SBF (r = 0.846, r = 0.716) and FBF versus CR (r = 0.700, r = 0.540). AR curves demonstrated higher SBF and FBF. CR remains almost constant in G1, G2, and G3. SBCI and FBCI increase significantly in G1, G2, and G3. Correlation between SBF and FBF and CR was strong for G1, G2, and very strong for G3. AR curves showed better correctability with SB and FB films.

Three-dimensional flow structure and patterns of bed shear stress in an evolving compound meander bend

Science.gov (United States)

Engel, Frank; Rhoads, Bruce L.

2016-01-01

Compound meander bends with multiple lobes of maximum curvature are common in actively evolving lowland rivers. Interaction among spatial patterns of mean flow, turbulence, bed morphology, bank failures and channel migration in compound bends is poorly understood. In this paper, acoustic Doppler current profiler (ADCP) measurements of the three-dimensional (3D) flow velocities in a compound bend are examined to evaluate the influence of channel curvature and hydrologic variability on the structure of flow within the bend. Flow structure at various flow stages is related to changes in bed morphology over the study timeframe. Increases in local curvature within the upstream lobe of the bend reduce outer bank velocities at morphologically significant flows, creating a region that protects the bank from high momentum flow and high bed shear stresses. The dimensionless radius of curvature in the upstream lobe is one-third less than that of the downstream lobe, with average bank erosion rates less than half of the erosion rates for the downstream lobe. Higher bank erosion rates within the downstream lobe correspond to the shift in a core of high velocity and bed shear stresses toward the outer bank as flow moves through the two lobes. These erosion patterns provide a mechanism for continued migration of the downstream lobe in the near future. Bed material size distributions within the bend correspond to spatial patterns of bed shear stress magnitudes, indicating that bed material sorting within the bend is governed by bed shear stress. Results suggest that patterns of flow, sediment entrainment, and planform evolution in compound meander bends are more complex than in simple meander bends. Moreover, interactions among local influences on the flow, such as woody debris, local topographic steering, and locally high curvature, tend to cause compound bends to evolve toward increasing planform complexity over time rather than stable configurations.

Recent developments in bend-insensitive and ultra-bend-insensitive fibers

Science.gov (United States)

Boivin, David; de Montmorillon, Louis-Anne; Provost, Lionel; Montaigne, Nelly; Gooijer, Frans; Aldea, Eugen; Jensma, Jaap; Sillard, Pierre

2010-02-01

Designed to overcome the limitations in case of extreme bending conditions, Bend- and Ultra-Bend-Insensitive Fibers (BIFs and UBIFs) appear as ideal solutions for use in FTTH networks and in components, pigtails or patch-cords for ever demanding applications such as military or sensing. Recently, however, questions have been raised concerning the Multi-Path-Interference (MPI) levels in these fibers. Indeed, they are potentially subject to interferences between the fundamental mode and the higher-order mode that is also bend resistant. This MPI is generated because of discrete discontinuities such as staples, bends and splices/connections that occur on distance scales that become comparable to the laser coherent length. In this paper, we will demonstrate the high MPI tolerance of all-solid single-trench-assisted BIFs and UBIFs. We will present the first comprehensive study combining theoretical and experimental points of view to quantify the impact of fusion splices on coherent MPI. To be complete, results for mechanical splices will also be reported. Finally, we will show how the single-trench- assisted concept combined with the versatile PCVD process allows to tightly control the distributions of fibers characteristics. Such controls are needed to massively produce BIFs and to meet the more stringent specifications of the UBIFs.

Degradation of aqueous phenol solutions by coaxial DBD reactor

Science.gov (United States)

Dojcinovic, B. P.; Manojlovic, D.; Roglic, G. M.; Obradovic, B. M.; Kuraica, M. M.; Puric, J.

2008-07-01

Solutions of 2-chlorophenol, 4-chlorophenol and 2,6-dichlorophenol in bidistilled and water from the river Danube were treated in plasma reactor. In this reactor, based on coaxial dielectric barrier discharge at atmospheric pressure, plasma is formed over a thin layer of treated water. After one pass through the reactor, starting chlorophenols concentration of 20 mg/l was diminished up to 95 %. Kinetics of the chlorophenols degradation was monitored by High Pressure Liquid Chromatography method (HPLC).

Thermal insulation system design and fabrication specification (nuclear) for the Clinch River Breeder Reactor plant

International Nuclear Information System (INIS)

1978-01-01

This specification defines the design, analysis, fabrication, testing, shipping, and quality requirements of the Insulation System for the Clinch River Breeder Reactor Plant (CRBRP), near Oak Ridge, Tennessee. The Insulation System includes all supports, convection barriers, jacketing, insulation, penetrations, fasteners, or other insulation support material or devices required to insulate the piping and equipment cryogenic and other special applications excluded. Site storage, handling and installation of the Insulation System are under the cognizance of the Purchaser

Investigation of cable deterioration in the containment building of the Savannah River Nuclear Reactor

International Nuclear Information System (INIS)

Gillen, K.T.; Clough, R.L.; Jones, L.H.

1982-08-01

This report describes an investigation of the deterioration of polyethylene and polyvinylchloride cable materials which occurred in the containment building of the Savannah River nuclear reactor located at Aiken, South Carolina. Radiation dosimetry and temperature mapping data of the containment area indicated that the maximum dose experienced by the cable materials was only 2.5 Mrad at an average operating temperature of 43 0 C. Considering this relatively moderate environment, the amount of material degradation seemed surprising. To understand these findings, an experimental program was performed on the commercial polyethylene and polyvinylchloride materials used at the plant to investigate their degradation behavior under combined γ-radiation and elevated temperature conditions. It is established that the material deterioration at the plant resulted from radiation-induced oxidation and that the degradation rate can be correlated with local levels of radiation intensity in the containment area

Shutdown of the River Water System at the Savannah River Site: Draft environmental impact statement

International Nuclear Information System (INIS)

1996-11-01

This environmental impact statement (EIS) evaluates alternative approaches to and environmental impacts of shutting down the River Water System at the Savannah River Site (SRS). Five production reactors were operated at the site.to support these facilities, the River Water System was constructed to provide cooling water to pass through heat exchangers to absorb heat from the reactor core in each of the five reactor areas (C, K, L, P, and R). The DOE Savannah River Strategic Plan directs the SRS to find ways to reduce operating costs and to determine what site infrastructure it must maintain and what infrastructure is surplus. The River Water System has been identified as a potential surplus facility. Three alternatives to reduce the River Water System operating costs are evaluated in this EIS. In addition to the No-Action Alternative, which consists of continuing to operate the River Water System, this EIS examines one alternative (the Preferred Alternative) to shut down and maintain the River Water System in a standby condition until DOE determines that a standby condition is no longer necessary, and one alternative to shut down and deactivate the River Water System. The document provides background information and introduces the River Water System at the SRS; sets forth the purpose and need for DOE action; describes the alternatives DOE is considering; describes the environment at the SRS and in the surrounding area potentially affected by the alternatives addressed and provides a detailed assessment of the potential environmental impacts of the alternatives; and identifies regulatory requirements and evaluates their applicability to the alternatives considered

Occipital bending in schizophrenia.

Science.gov (United States)

Maller, Jerome J; Anderson, Rodney J; Thomson, Richard H; Daskalakis, Zafiris J; Rosenfeld, Jeffrey V; Fitzgerald, Paul B

2017-01-01

To investigate the prevalence of occipital bending (an occipital lobe crossing or twisting across the midline) in subjects with schizophrenia and matched healthy controls. Occipital bending prevalence was investigated in 37 patients with schizophrenia and 44 healthy controls. Ratings showed that prevalence was nearly three times higher among schizophrenia patients (13/37 [35.1%]) than in control subjects (6/44 [13.6%]). Furthermore, those with schizophrenia had greater normalized gray matter volume but less white matter volume and had larger brain-to-cranial ratio. The results suggest that occipital bending is more prevalent among schizophrenia patients than healthy subjects and that schizophrenia patients have different gray matter-white matter proportions. Although the cause and clinical ramifications of occipital bending are unclear, the results infer that occipital bending may be a marker of psychiatric illness.

Band-Bending of Ga-Polar GaN Interfaced with Al2O3 through Ultraviolet/Ozone Treatment.

Science.gov (United States)

Kim, Kwangeun; Ryu, Jae Ha; Kim, Jisoo; Cho, Sang June; Liu, Dong; Park, Jeongpil; Lee, In-Kyu; Moody, Baxter; Zhou, Weidong; Albrecht, John; Ma, Zhenqiang

2017-05-24

Understanding the band bending at the interface of GaN/dielectric under different surface treatment conditions is critically important for device design, device performance, and device reliability. The effects of ultraviolet/ozone (UV/O 3 ) treatment of the GaN surface on the energy band bending of atomic-layer-deposition (ALD) Al 2 O 3 coated Ga-polar GaN were studied. The UV/O 3 treatment and post-ALD anneal can be used to effectively vary the band bending, the valence band offset, conduction band offset, and the interface dipole at the Al 2 O 3 /GaN interfaces. The UV/O 3 treatment increases the surface energy of the Ga-polar GaN, improves the uniformity of Al 2 O 3 deposition, and changes the amount of trapped charges in the ALD layer. The positively charged surface states formed by the UV/O 3 treatment-induced surface factors externally screen the effect of polarization charges in the GaN, in effect, determining the eventual energy band bending at the Al 2 O 3 /GaN interfaces. An optimal UV/O 3 treatment condition also exists for realizing the "best" interface conditions. The study of UV/O 3 treatment effect on the band alignments at the dielectric/III-nitride interfaces will be valuable for applications of transistors, light-emitting diodes, and photovoltaics.

Variability of Darcian Flux in the Hyporheic Zone at a Natural Channel Bend

Directory of Open Access Journals (Sweden)

Shaofeng Xu

2017-02-01

Full Text Available Channel bends are one of the most important characteristic features of natural streams. These bends often create the conditions for a hyporheic zone, which has been recognized as a critical component of stream ecosystems. The streambed vertical hydraulic conductivity (Kv, vertical hydraulic gradient (VHG and Darcian flux (DF in the hyporheic zone were estimated at 61 locations along a channel bend of the Beiluo River during July 2015 and January 2016. All the streambed attributes showed great spatial variability along the channel bend. Both upward fluxes and downward fluxes occurred during the two test periods, most of studied stream sections were controlled by downwelling, indicating stream water discharge into the subsurface. The average downward flux was higher at the downstream side than at the upstream side of the channel bend, especially in July 2015. The distribution of streambed sediment grain size has a significant influence on the variability of Kv; high percentages of silt and clay sediments generally lead to low Kv values. Higher Kv at the depositional left bank at the upstream site shifted toward the erosional right bank at the downstream site, with Kv values positively correlated with the water depth. This study suggested that the variabilities of Kv and VHG were influenced by the stream geomorphology and that the distribution of Kv was inversely related, to a certain extent, to the distribution of VHG across the channel bend. Kv and VHG were found to have opposite effects on the DF, and the close relationship between Kv and DF indicated that the water fluxes were mainly controlled by Kv.

Elastic properties of nanolaminar Cr_2AlC films and beams determined by in-situ scanning electron microscope bending tests

International Nuclear Information System (INIS)

Grieseler, Rolf; Theska, Felix; Stürzel, Thomas; Hähnlein, Bernd; Stubenrauch, Mike; Hopfeld, Marcus; Kups, Thomas; Pezoldt, Jörg; Schaaf, Peter

2016-01-01

The mechanical properties of Cr_2AlC MAX phase structures were investigated by in-situ bending tests. Freestanding structures such as cantilevers and doubly clamped beams of Cr_2AlC were produced. The structures exhibit a Young's modulus of 184 GPa which is close to the value obtained by vibrational measurements. The in-situ bending test allows the determination of the mechanical properties with a lower variance of the measurement results compared to the vibrational measurement. The results are a good starting point for the development of microelectromechanical structures based on MAX phases. - Highlights: • Cr_2AlC were produced by deposition multilayers and subsequent rapid annealing. • Freestanding doubly clamped beams and cantilevers of Cr_2AlC were prepared. • A finite elements model was implemented showing the displacement of the structure. • In-situ bending test at doubly clamped beams and cantilevers were performed. • An in-situ bending test is a valid approach to determine mechanical properties.

Upgrade of the Department of Energy's Savannah River Site's reactor operations and maintenance procedures

International Nuclear Information System (INIS)

Walsh, T.E.

1991-01-01

This paper describes the program in progress at the Savannah River Site (SRS) to upgrade the existing reactor operating and maintenance procedures to current commercial nuclear industry standards. In order to meet this goal, the following elements were established: administrative procedures to govern the upgrade process, tracking system to provide status and accountability; and procedure writing guides. The goal is to establish a benchmark of excellence by which other Department of Energy (DOE) sites will measure themselves. The above three elements are addressed in detail in this paper

PARR-2: reactor description and experiments

International Nuclear Information System (INIS)

Wyne, M.F.; Meghji, J.H.

1990-12-01

PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

Comparison of ring compression testing to three point bend testing for unirradiated ZIRLO cladding

Energy Technology Data Exchange (ETDEWEB)

None, None

2015-04-01

Safe shipment and storage of nuclear reactor discharged fuel requires an understanding of how the fuel may perform under the various conditions that can be encountered. One specific focus of concern is performance during a shipment drop accident. Tests at Savannah River National Laboratory (SRNL) are being performed to characterize the properties of fuel clad relative to a mechanical accident condition such as a container drop. Unirradiated ZIRLO tubing samples have been charged with a range of hydride levels to simulate actual fuel rod levels. Samples of the hydrogen charged tubes were exposed to a radial hydride growth treatment (RHGT) consisting of heating to 400°C, applying initial hoop stresses of 90 to 170 MPa with controlled cooling and producing hydride precipitates. Initial samples have been tested using both a) ring compression test (RCT) which is shown to be sensitive to radial hydride and b) three-point bend tests which are less sensitive to radial hydride effects. Hydrides are generated in Zirconium based fuel cladding as a result of coolant (water) oxidation of the clad, hydrogen release, and a portion of the released (nascent) hydrogen absorbed into the clad and eventually exceeding the hydrogen solubility limit. The orientation of the hydrides relative to the subsequent normal and accident strains has a significant impact on the failure susceptability. In this study the impacts of stress, temperature and hydrogen levels are evaluated in reference to the propensity for hydride reorientation from the circumferential to the radial orientation. In addition the effects of radial hydrides on the Quasi Ductile Brittle Transition Temperature (DBTT) were measured. The results suggest that a) the severity of the radial hydride impact is related to the hydrogen level-peak temperature combination (for example at a peak drying temperature of 400°C; 800 PPM hydrogen has less of an impact/ less radial hydride fraction than 200 PPM hydrogen for the same thermal

Reactor BR2: Introduction

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. A safety audit was conduced by the IAEA, the conclusions of which demonstrated the excellent performance of the plant in terms of operational safety. In 1999, the CALLISTO facility was extensively used for various programmes involving LWR pressure vessel materials, IASCC of LWR structural materials, fusion reactor materials and martensic steels for use in ADS systems. In 1999, BR2's commercial programmes were further developed

An analysis of a pipe bend subjected to in-plane loads

International Nuclear Information System (INIS)

Hellen, T.K.

1979-01-01

This report describes a set of finite element analyses conducted on a pipe bend subjected to in-plane loads. The pipe is thin-walled, and two types of finite element, shells and solid bricks, are compared elastically. An alternative semi-analytical technique has also been used and experimental results are available, all of which show good correlative agreement. The use of suitable mesh refinement and order of numerical integration is examined. Finally, the solid elements are used to follow a loading sequence incorporating elasto-plastic behaviour as conducted by experiment. This work is an updated version of that used for the CEC benchmark calculations for the Fast Reactor Codes and Standards Working Group, Activity No 2, on Structural Analysis. (author)

Historical Channel Adjustment and Estimates of Selected Hydraulic Values in the Lower Sabine River and Lower Brazos River Basins, Texas and Louisiana

Science.gov (United States)

Heitmuller, Franklin T.; Greene, Lauren E.

2009-01-01

The U.S. Geological Survey, in cooperation with the Texas Water Development Board, evaluated historical channel adjustment and estimated selected hydraulic values at U.S. Geological Survey streamflow-gaging stations in the lower Sabine River Basin in Texas and Louisiana and lower Brazos River Basin in Texas to support geomorphic assessments of the Texas Instream Flow Program. Channel attributes including cross-section geometry, slope, and planform change were evaluated to learn how each river's morphology changed over the years in response to natural and anthropogenic disturbances. Historical and contemporary cross-sectional channel geometries at several gaging stations on each river were compared, planform changes were assessed, and hydraulic values were estimated including mean flow velocity, bed shear stress, Froude numbers, and hydraulic depth. The primary sources of historical channel morphology information were U.S. Geological Survey hard-copy discharge-measurement field notes. Additional analyses were done using computations of selected flow hydraulics, comparisons of historical and contemporary aerial photographs, comparisons of historical and contemporary ground photographs, evaluations of how frequently stage-discharge rating curves were updated, reviews of stage-discharge relations for field measurements, and considerations of bridge and reservoir construction activities. Based on historical cross sections at three gaging stations downstream from Toledo Bend Reservoir, the lower Sabine River is relatively stable, but is subject to substantial temporary scour-and-fill processes during floods. Exceptions to this characterization of relative stability include an episode of channel aggradation at the Sabine River near Bon Wier, Texas, during the 1930s, and about 2 to 3 feet of channel incision at the Sabine River near Burkeville, Texas, since the late 1950s. The Brazos River, at gaging stations downstream from Waco, Texas, has adjusted to a combination of

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

International Nuclear Information System (INIS)

Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

1990-01-01

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2001-01-01

The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2001-04-01

The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

Bank pull or bar push: What drives scroll-bar formation in meandering rivers?

NARCIS (Netherlands)

van de Lageweg, W. I.; van Dijk, W. M.; Baar, A. W.; Rutten, J.; Kleinhans, M. G.

2014-01-01

One of the most striking features of meandering rivers are quasi-regular ridges of the point bar, evidence of a pulsed lateral migration of meander bends. Scroll bars formed on the inner bend are preserved on the point-bar surface as a series of ridges as meanders migrate, and in the subsurface of

Standard test methods for bend testing of material for ductility

CERN Document Server

American Society for Testing and Materials. Philadelphia

2009-01-01

1.1 These test methods cover bend testing for ductility of materials. Included in the procedures are four conditions of constraint on the bent portion of the specimen; a guided-bend test using a mandrel or plunger of defined dimensions to force the mid-length of the specimen between two supports separated by a defined space; a semi-guided bend test in which the specimen is bent, while in contact with a mandrel, through a specified angle or to a specified inside radius (r) of curvature, measured while under the bending force; a free-bend test in which the ends of the specimen are brought toward each other, but in which no transverse force is applied to the bend itself and there is no contact of the concave inside surface of the bend with other material; a bend and flatten test, in which a transverse force is applied to the bend such that the legs make contact with each other over the length of the specimen. 1.2 After bending, the convex surface of the bend is examined for evidence of a crack or surface irregu...

Flow visualization study of two phase flow in a single bend outlet feeder pipe and horizontal annulus of outlet end-fitting of a CANDU reactor

International Nuclear Information System (INIS)

Supa-Amornkul, S.; Lister, D.H.; Steward, F.R.

2005-01-01

'Full text:' In CANDU-6 reactors, the pressurized high-temperature coolant flows through 380 fuel channels passing horizontally through the core. Each end of a fuel channel has a stainless steel annular end-fitting connected to a carbon steel feeder pipe. The outlet coolant, which is at 310 o C with up to 0.30 steam voidage, turns through 90 o as it passes from flow in the annular end-fitting to pipe flow in the feeder via a Grayloc connector. Since 1996, several CANDU stations have reported excessive corrosion of their outlet feeder pipes; especially between the first metre, which consisted of single or double bends. Early studies related the attack to the hydrodynamics of the coolant and verified that it was a type of flow accelerated corrosion. In order to understand the hydrodynamics of the coolant in the outlet feeders by flow-visualization, a full-scale transparent test section simulating the geometry and orientation of an outlet feeder bend with its upstream annular end-fitting were fabricated. The feeder consisted of a 54 mm inside diameter acrylic pipe with a 73 o bend, connecting to an acrylic simulation of a Grayloc flanged fitting and annular end-fitting. The annular end-fitting consisted of an inner pipe, 110 mm outer diameter, and an outer pipe, 150 mm inner diameter, both 190.7 cm long in length. The tests were performed with water and air at atmospheric pressure and room temperature. The maximum water volumetric flow rate was 19 L/s and the volume fraction of air varied from 0.05 to 0.56. The phase distributions within the feeder pipe and along the length of the annulus were investigated with a digital video recorder. Size, concentration and velocity of the air bubbles at particular locations were studied with a high-speed digital still camera and a high-speed digital video camera. Phase distributions and variations in bubble size with velocity were determined. Particular attention was paid to the flow pattern at the inside of the bend, where a CFD

Flow visualization study of two phase flow in a single bend outlet feeder pipe and horizontal annulus of outlet end-fitting of a CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Supa-Amornkul, S.; Lister, D.H.; Steward, F.R. [Univ. of New Brunswick, Fredericton, New Brunswick (Canada)]. E-mail: h796e@unb.ca; dlister@unb.ca; fsteward@unb.ca

2005-07-01

'Full text:' In CANDU-6 reactors, the pressurized high-temperature coolant flows through 380 fuel channels passing horizontally through the core. Each end of a fuel channel has a stainless steel annular end-fitting connected to a carbon steel feeder pipe. The outlet coolant, which is at 310{sup o}C with up to 0.30 steam voidage, turns through 90{sup o} as it passes from flow in the annular end-fitting to pipe flow in the feeder via a Grayloc connector. Since 1996, several CANDU stations have reported excessive corrosion of their outlet feeder pipes; especially between the first metre, which consisted of single or double bends. Early studies related the attack to the hydrodynamics of the coolant and verified that it was a type of flow accelerated corrosion. In order to understand the hydrodynamics of the coolant in the outlet feeders by flow-visualization, a full-scale transparent test section simulating the geometry and orientation of an outlet feeder bend with its upstream annular end-fitting were fabricated. The feeder consisted of a 54 mm inside diameter acrylic pipe with a 73{sup o} bend, connecting to an acrylic simulation of a Grayloc flanged fitting and annular end-fitting. The annular end-fitting consisted of an inner pipe, 110 mm outer diameter, and an outer pipe, 150 mm inner diameter, both 190.7 cm long in length. The tests were performed with water and air at atmospheric pressure and room temperature. The maximum water volumetric flow rate was 19 L/s and the volume fraction of air varied from 0.05 to 0.56. The phase distributions within the feeder pipe and along the length of the annulus were investigated with a digital video recorder. Size, concentration and velocity of the air bubbles at particular locations were studied with a high-speed digital still camera and a high-speed digital video camera. Phase distributions and variations in bubble size with velocity were determined. Particular attention was paid to the flow pattern at the inside

Investigation of cable deterioration in the containment building of the Savannah River Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Gillen, K.T.; Clough, R.L.; Jones, L.H.

1982-08-01

This report describes an investigation of the deterioration of polyethylene and polyvinylchloride cable materials which occurred in the containment building of the Savannah River nuclear reactor located at Aiken, South Carolina. Radiation dosimetry and temperature mapping data of the containment area indicated that the maximum dose experienced by the cable materials was only 2.5 Mrad at an average operating temperature of 43/sup 0/C. Considering this relatively moderate environment, the amount of material degradation seemed surprising. To understand these findings, an experimental program was performed on the commercial polyethylene and polyvinylchloride materials used at the plant to investigate their degradation behavior under combined ..gamma..-radiation and elevated temperature conditions. It is established that the material deterioration at the plant resulted from radiation-induced oxidation and that the degradation rate can be correlated with local levels of radiation intensity in the containment area.

Safety features of TR-2 reactor

International Nuclear Information System (INIS)

Tuerker, T.

2001-01-01

TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

Bending Strength of EN AC-44200 – Al2O3 Composites at Elevated Temperatures

Directory of Open Access Journals (Sweden)

Kurzawa A.

2017-03-01

Full Text Available The paper presents results of bend tests at elevated temperatures of aluminium alloy EN AC-44200 (AlSi12 based composite materials reinforced with aluminium oxide particles. The examined materials were manufactured by squeeze casting. Preforms made of Al2O3 particles, with volumetric fraction 10, 20, 30 and 40 vol.% of particles joined with sodium silicate bridges were used as reinforcement. The preforms were characterised by open porosity ensuring proper infiltration with the EN AC-44200 (AlSi12 liquid alloy. The largest bending strength was found for the materials containing 40 vol.% of reinforcing ceramic particles, tested at ambient temperature. At increased test temperature, bending strength Rg of composites decreased in average by 30 to 50 MPa per 100°C of temperature increase. Temperature increase did not significantly affect cracking of the materials. Cracks propagated mainly along the interfaces particle/matrix, with no effect of the particles falling-out from fracture surfaces. Direction of cracking can be affected by a small number of agglomerations of particles or of non-reacted binder. In the composites, the particles strongly restrict plastic deformation of the alloy, which leads to creation of brittle fractures. At elevated temperatures, however mainly at 200 and 300°C, larger numbers of broken, fragmented particles was observed in the vicinity of cracks. Fragmentation of particles occurred mainly at tensioned side of the bended specimens, in the materials with smaller fraction of Al2O3 reinforcement, i.e. 10 and 20 vol.%.

Don't Fence Me In: Free Meanders in a Confined River Valley

Science.gov (United States)

Eke, E. C.; Wilcock, P. R.

2015-12-01

The interaction between meandering river channels and inerodible valley walls provides a useful test of our ability to understand meander dynamics. In some cases, river meanders confined between valley walls display distinctive angular bends in a dynamic equilibrium such that their size and shape persist as the meander migrates. In other cases, meander geometry is more varied and changes as the meander migrates. The ratio of channel to valley width has been identified as a useful parameter for defining confined meanders, but is not sufficient to distinguish cases in which sharp angular bends are able to migrate with little change in geometry. Here, we examine the effect of water and sediment supply on the geometry of confined rivers in order to identify conditions under which meander geometry reaches a persistent dynamic equilibrium. Because channel width and meander geometry are closely related, we use a numerical meander model that allows for independent migration of both banks, thereby allowing channel width to vary in space and time. We hypothesize that confined meanders with persistent angular bends have smaller transport rates of bed material and that their migration is driven by erosion of the cutbank (bank-pull migration). When bed material supply is sufficiently large that point bar deposition drives meander migration (bar-push migration), confined meander bends have a larger radius of curvature and a geometry that varies as the meander migrates. We test this hypothesis using historical patterns of confined meander migration for rivers with different rates of sediment supply and bed material transport. Interpretation of the meander migration pattern is provided by the free-width meander migration model.

Bending the law: tidal bending and its effects on ice viscosity and flow

Science.gov (United States)

Rosier, S.; Gudmundsson, G. H.

2017-12-01

Many ice shelves are subject to strong ocean tides and, in order to accommodate this vertical motion, the ice must bend within the grounding zone. This tidal bending generates large stresses within the ice, changing its effective viscosity. For a confined ice shelf, this is particularly relevant because the tidal bending stresses occur along the sidewalls, which play an important role in the overall flow regime of the ice shelf. Hence, tidal bending stresses will affect both the mean and time-varying components of ice shelf flow. GPS measurements reveal strong variations in horizontal ice shelf velocities at a variety of tidal frequencies. We show, using full-Stokes viscoelastic modelling, that inclusion of tidal bending within the model accounts for much of the observed tidal modulation of horizontal ice shelf flow. Furthermore, our model shows that in the absence of a vertical tidal forcing, the mean flow of the ice shelf is reduced considerably.

Effect of contouring on bending structural stiffness and bending strength of the 3.5 titanium SOP implant.

Science.gov (United States)

Rutherford, Scott; Ness, Malcolm G

2012-11-01

To compare the bending structural stiffness (BSS) and bending strength (BS) of the 3.5 titanium (Ti) string of pearls (SOP) plate and the 3.5 316LVM stainless steel SOP plate; and the effect of contouring on the BSS and BS of the 3.5 Ti SOP plate. In vitro experimental static 4-point bending materials testing. Twenty-five 3.5 mm Ti and five 3.5 mm 316LVM stainless steel SOP locking bone plates. Each plate was tested in 4-point bending until 10 mm of displacement was achieved. BSS and BS were then calculated for each plate. A 2-sample t-test was used to compare the mean BSS and BS of the different groups. The 3.5 Ti SOP plate had lower mean BSS (0.00263 Nm(2) ) but similar mean BS (12.8 Nm) when compared to the 3.5 316LVM SOP (0.00402 Nm(2) , 13.0 Nm). Prebending the 3.5 Ti SOP diminished its mean BSS (0.00224 Nm(2) ) and mean BS (9.4 Nm) when compared to the Ti control. Pretwisting the 3.5 Ti SOP increased its mean BSS (0.00273 Nm(2) ) but decreased its mean BS (12.4 Nm) when compared to the Ti control. The 3.5 Ti SOP is less stiff but of similar strength to the 3.5 316LVM stainless steel SOP. Prebending the Ti SOP significantly lowers its stiffness and strength. Pretwisting the SOP actually increases its stiffness but slightly lowers its strength. © Copyright 2012 by The American College of Veterinary Surgeons.

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2002-01-01

The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2002-04-01

The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

A New Kind of Bend Sensor

Institute of Scientific and Technical Information of China (English)

无

2000-01-01

A new kind of bend sensor is introduced.It can be used to detect the bend angle of an object or inclination between two objects.It has characteristics of small size, lightweight, high reliability, fine flexibility and plasticity.When this bend sensor is used with a proper converting circuit, it can implement dynamic measuring the bend angle of an object conveniently.The application of the bend sensor in dataglove is also described.

Reprocessing fuel from the Southwest Experimental Fast Oxide Reactor at the Savannah River Plant

International Nuclear Information System (INIS)

Gray, L.W.; Campbell, T.G.

1985-11-01

The irradiated fuel, reject fuel tubes, and fuel fabrication scrap from the Southwest Experimental Fast Oxide Reactor (SEFOR) were transferred to the Savannah River Plant (SRP) for uranium and plutonium recovery. The unirradiated material was declad and dissolved at SRP; dissolution was accomplished in concentrated nitric acid without the addition of fluoride. The irradiated fuel was declad at Atomics International and repacked in aluminum. The fuel and aluminum cans were dissolved at SRP using nitric acid catalyzed by mercuric nitrate. As this fuel was dissolved in nongeometrically favorable tanks, boron was used as a soluble neutron poison

ELECTRICAL RESISTANCE HEATING OF SOILS AT C-REACTOR AT THE SAVANNAH RIVER SITE

International Nuclear Information System (INIS)

Blundy, R; Michael Morgenstern, M; Joseph Amari, J; Annamarie MacMurray, A; Mark Farrar, M; Terry Killeen, T

2007-01-01

Chlorinated solvent contamination of soils and groundwater is an endemic problem at the Savannah River Site (SRS), and originated as by-products from the nuclear materials manufacturing process. Five nuclear reactors at the SRS produced special nuclear materials for the nation's defense program throughout the cold war era. An important step in the process was thorough degreasing of the fuel and target assemblies prior to irradiation. Discharges from this degreasing process resulted in significant groundwater contamination that would continue well into the future unless a soil remediation action was performed. The largest reactor contamination plume originated from C-Reactor and an interim action was selected in 2004 to remove the residual trichloroethylene (TCE) source material by electrical resistance heating (ERH) technology. This would be followed by monitoring to determine the rate of decrease in concentration in the contaminant plume. Because of the existence of numerous chlorinated solvent sources around SRS, it was elected to generate in-house expertise in the design and operation of ERH, together with the construction of a portable ERH/SVE system that could be deployed at multiple locations around the site. This paper describes the waste unit characteristics, the ERH system design and operation, together with extensive data accumulated from the first deployment adjacent to the C-Reactor building. The installation heated the vadose zone down to 62 feet bgs over a 60 day period during the summer of 2006 and raised soil temperatures to over 200 F. A total of 730 lbs of trichloroethylene (TCE) were removed over this period, and subsequent sampling indicated a removal efficiency of 99.4%

Study of transport and micro-structural properties of magnesium di-boride strand under react and bend mode and bend and react mode

International Nuclear Information System (INIS)

Kundu, Ananya; Das, Subrat Kumar; Bano, Anees; Pradhan, Subrata

2015-01-01

I-V characterization of commercial multi-filamentary Magnesium Di-Boride (MgB 2 ) wire of diameter 0.83 mm were studied in cryocooler based self-field characterization system under both react and bent mode and bent and react mode for a range of temperature 6 K - 25 K. This study is of practical technical relevance where the heat treatment of the superconducting wire makes the sample less flexible for winding in magnet and in other applications. There are limited reported data, available on degradation of MgB 2 wire with bending induced strain in react and wind and wind and react method. In the present work the bending diameter were varied from 80 mm to 20 mm in the interval of 10 mm change of bending diameter and for each case critical current (Ic) of the strand is measured for the above range of temperature. An ETP copper made customized sample holder for mounting the MgB 2 strand was fabricated and is thermally anchored to the cooling stage of the cryocooler. It is seen from the experimental data that in react and bent mode the critical current degrades from 105 A to 87 A corresponding to bending diameter of 80 mm and 20 mm respectively. The corresponding bending strain was analytically estimated and compared with the simulation result. It is also observed that in react and bent mode, the degradation of the transport property of the strand is less as compared to react and bent mode. For bent and react mode in the same sample, the critical current (Ic) was measured to be ∼145 A at 15 K for bending diameter of 20 mm. Apart from studying the bending induced strain on MgB 2 strand, the tensile test of the strand at RT was carried out. The electrical characterizations of the samples were accompanied by the microstructure analyses of the bent strand to examine the bending induced degradation in the grain structure of the strand. All these experimental findings are expected to be used as input to fabricate prototype MgB 2 based magnet. (author)

Influence of P-Reactor operation on the aquatic ecology of Par Pond: a literature review

International Nuclear Information System (INIS)

Wilde, E.W.; Tilly, L.J.

1985-02-01

Par Pond is a 1012 hectare reservoir that was constructed in 1958 to provide cooling water for Savannah River nuclear reactors. The purpose of this report is to summarize all known studies on the Par Pond system and point out demonstrable or probable effects that can be correlated with reactor operations. Reactor operation effects the Par Pond ecosystem through: (1) pumping, (2) thermal alteration, and (3) the addition of Savannah River makeup water. The influence of each of these factors is discussed. 108 references, 24 figures, 34 tables. (MF)

Distribution of Hanford reactor produced radionuclides in the marine environment, 1961-73

International Nuclear Information System (INIS)

Seymour, A.H.

1980-01-01

At Hanford (U.S.A.), the plutonium-producing reactors were in operation during 1944-1971. The period of maximum reactor operation was 1955-1965, when eight reactors were in operation. The reactor deactivation programme began in 1965 and the last reactor was deactivated in 1971. All these reactors were cooled by Columbia River water which passed through the reactors and then was discharged to the river and ultimately to the North Pacific. The Laboratory of Radiation Ecology (LRE) of the University of Washington started an environmental survey programme in 1965 and continued it upto 1973 i.e. even after the last plutonium producing reactor was deactivated. The programme objectives were: (1) to find the geographical distribution and concentration of Hanford produced radionuclides in water, sediments and biota of the marine environment, (2) to relate the operation of the Hanford reactors during the period of deactivation to the concentration of radionuclides in marine organisms, and (3) to observe the rate at which the marine organisms cleansed themselves of 65 Zn after the primary source had been removed. An account of the programme and highlights of the observations are reported. Most of the radioactivity entering the river water and marine organisms was due to 51 Cr, 65 Zn and 32 P of which 65 Zn was found to be the most abundant radionuclide in the biological samples. The rate of radioactivity from the river water entering into the Ocean was about 1000 curies per day and it did not produce any observable effects on populations of marine organisms. The internal dose to man from 65 Zn via seafoods was only a small fraction of the permissible dose for individual members of the population. (M.G.B.)

Scram and nonlinear reactor system seismic analysis for a liquid metal fast reactor

International Nuclear Information System (INIS)

Morrone, A.; Brussalis, W.G.

1975-01-01

The paper presents the analysis and results for a LMFBR system which was analyzed for both scram times and seismic responses such as bending moments, accelerations and forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node (translational and rotational). The model was developed to incorporate as many reactor components as possible without exceeding computer limitations. It consists of 12 reactor components with a total of 71 nodes, 69 beam and pin-jointed elements and 27 gap elements. The gap elements were defined by their clearances, impact spring constants and impact damping constants based on a 50% coefficient of restitution. The horizontal excitation input to the model was the response of the containment building at the location of the reactor vessel supports. It consists of a ten seconds Safe Shutdown Earthquake acceleration-time history at 0.005 seconds intervals and with a maximum acceleration of 0.408 g. The analysis was performed with two Westinghouse special purpose computer programs. The first program calculated the reactor system seismic responses and stored the impact forces on tape. The impact forces on the control rod driveline were converted into vertical frictional forces by multiplying them by a coefficient of friction, and then used by the second program for the scram time determination. The results give time history plots of various seismic responses, and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about 4 times longer than that calculated without the earthquake. The bending moment and shear force responses were used as input for the structural analysis (stresses, deflections, fatigue) of the various components, in combination with the other applicable loading conditions. (orig./HP) [de

BR2 Reactor: Introduction

International Nuclear Information System (INIS)

Moons, F.

2007-01-01

The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3 He, 6 Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

The bankfull hydraulic geometry of evolving meander bends

Science.gov (United States)

Monegaglia, F.; Tubino, M.; Zolezzi, G.

2017-12-01

Changes in the bankfull hydraulic geometry of meandering rivers associated with meander growth from incipient meandering to cutoffs have seldom been analysed in detail. Such information is also needed by meander morphodynamic models, most of which simulate the evolution of bankfull channel geometry by simply accounting for channel slope reduction inversely proportional to elongation, while changes in bankfull channel width are often neglected or, when they are considered, they are not consistent with the few available observations. To address these gaps we first perform an extensive, systematic, bend-scale evolutionary analysis of bankfull channel widths in several large meandering rivers in the Amazon basin, over a three decades time period, from remotely sensed field data. The analysis consistently show a slight decreasing trend of the bankfull channel width during the planform evolution towards cutoff. Furthermore, we develop a physically based model for the evolution of bankfull channel geometry during the planform development of meandering rivers. The model is based on the conservation of sediment discharge. An integrated one-dimensional Exner equation that accounts for meander elongation, sediment supply conservation and sediment income from the channel banks, allows us to predict the evolution of the channel slope. The evolution of the channel width is modeled through a threshold equation. The model correctly predicts the slight variability of channel width during meander development and a gentler reduction of the channel slope, which is mitigated by the conservation of sediment supply. The bankfull geometry of highly dynamic meandering rivers is predicted to be elongation-dominated, while the one related to slowly evolving meandering rivers is sediment supply-dominated. Finally, we discuss the implications of the proposed modeling framework in terms of planform structure, meander shape and morphodynamic influence.

Application of the SQUG-GIP to the seismic upgrade program of the Savannah River reactors

International Nuclear Information System (INIS)

Antaki, G.A.

1991-01-01

In August 1991, the Savannah River Site (SRS) seismic evaluation program using the Generic Implementation Procedure (GIP) celebrated its third anniversary-a respectable age for such a new methodology. During these three years, the GIP, developed for the commercial nuclear industry's Seismic Qualification Utility Group (SQUG), had evolved through Revision 01, Revision 1, Revision 2 and a Revision 2 open-quotes updateclose quotes which is currently in the works. This evolution is not surprising for such an important, and in many ways pioneering, document. The various revisions were anticipated at SRS, and the program adjusted accordingly. The verification of seismic adequacy of equipment at the SRS nuclear reactors has been outlined in previous publications. The purpose of this paper is to relate the more practical and managerial aspects of our relatively mature SQUG-GIP implementation program, which will hopefully prove useful to future users of the GIP. This report is divided into four sections, which follow the normal flow of work under GIP: (1) Program Prerequisites; (2) Definition of Scope; (3) Equipment Evaluations; and (4) Resolution of Outliers

CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 1, 2, AND 3 OF CRYSTAL RIVER UNIT 3

International Nuclear Information System (INIS)

Wright, Kenneth D.

1997-01-01

The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 1, 2, and 3 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies

Longitudinal-bending mode micromotor using multilayer piezoelectric actuator.

Science.gov (United States)

Yao, K; Koc, B; Uchino, K

2001-07-01

Longitudinal-bending mode ultrasonic motors with a diameter of 3 mm were fabricated using stacked multilayer piezoelectric actuators, which were self-developed from hard lead zirconate titanate (PZT) ceramic. A bending vibration was converted from a longitudinal vibration with a longitudinal-bending coupler. The motors could be bidirectionally operated by changing driving frequency. Their starting and braking torque were analyzed based on the transient velocity response. With a load of moment of inertia 2.5 x 10(-7) kgm2, the motor showed a maximum starting torque of 127.5 microNm. The braking torque proved to be a constant independent on the motor's driving conditions and was roughly equivalent to the maximum starting torque achievable with our micromotors.

Weld repair of helium degraded reactor vessel material

International Nuclear Information System (INIS)

Kanne, W.R. Jr.; Lohmeier, D.A.; Louthan, M.R. Jr.; Rankin, D.T.; Franco-Ferreira, E.A.; Bruck, G.J.; Madeyski, A.; Shogan, R.P.; Lessmann, G.G.

1990-01-01

Welding methods for modification or repair of irradiated nuclear reactor vessels are being evaluated at the Savannah River Site. A low-penetration weld overlay technique has been developed to minimize the adverse effects of irradiation induced helium on the weldability of metals and alloys. This technique was successfully applied to Type 304 stainless steel test plates that contained 3 to 220 appm helium from tritium decay. Conventional welding practices caused significant cracking and degradation in the test plates. Optical microscopy of weld surfaces and cross sections showed that large surface toe cracks formed around conventional welds in the test plates but did not form around overlay welds. Scattered incipient underbead cracks (grain boundary separations) were associated with both conventional and overlay test welds. Tensile and bend tests were used to assess the effect of base metal helium content on the mechanical integrity of the low-penetration overlay welds. The axis of tensile specimens was perpendicular to the weld-base metal interface. Tensile specimens were machined after studs were resistance welded to overlay surfaces

Optimum Pathways of Fish Spawning Migrations in Rivers

Science.gov (United States)

McElroy, B. J.; Jacobson, R. B.; Delonay, A.

2010-12-01

Many fish species migrate large distances upstream in rivers to spawn. These migrations require energetic expenditures that are inversely related to fecundity of spawners. Here we present the theory necessary to quantify relative energetic requirements of upstream migration pathways and then test the hypothesis that least-cost paths are taken by the federally endangered pallid sturgeon (Scaphyrhyncus Albus), a benthic rheophile, in the lower Missouri River, USA. Total work done by a fish through a migratory path is proportional to the size of the fish, the total drag on the fish, and the distance traversed. Normalizing by the work required to remain stationary at the beginning of a path, relative work expenditure at each point of the path is found to be the cube of the ratio of the velocity along the path to the velocity at the start of the path. This is the velocity of the fish relative to the river flow. A least-cost migratory pathway can be determined from the velocity field in a reach as the path that minimizes a fish's relative work expenditure. We combine location data from pallid sturgeon implanted with telemetric tags and pressure-sensitive data storage tags with depth and velocity data collected with an acoustic Doppler profiler. During spring 2010 individual sturgeon were closely followed as they migrated up the Missouri River to spawn. These show that, within a small margin, pallid sturgeon in the lower Missouri River select least-cost paths as they swim upstream (typical velocities near 1.0 - 1.2 m/s). Within the range of collected data, it is also seen that many alternative paths not selected for migration are two orders of magnitude more energetically expensive (typical velocities near 2.0 - 2.5 m/s). In general these sturgeon migrated along the inner banks of bends avoiding high velocities in the thalweg, crossing the channel where the thalweg crosses in the opposite direction in order to proceed up the inner bank of subsequent bends. Overall, these

A 2-DOF microstructure-dependent model for the coupled torsion/bending instability of rotational nanoscanner

Science.gov (United States)

Keivani, M.; Abadian, N.; Koochi, A.; Mokhtari, J.; Abadyan, M.

2016-10-01

It has been well established that the physical performance of nanodevices might be affected by the microstructure. Herein, a two-degree-of-freedom model base on the modified couple stress theory is developed to incorporate the impact of microstructure in the torsion/bending coupled instability of rotational nanoscanner. Effect of microstructure dependency on the instability parameters is determined as a function of the microstructure parameter, bending/torsion coupling ratio, van der Waals force parameter and geometrical dimensions. It is found that the bending/torsion coupling substantially affects the stable behavior of the scanners especially those with long rotational beam elements. Impact of microstructure on instability voltage of the nanoscanner depends on coupling ratio and the conquering bending mode over torsion mode. This effect is more highlighted for higher values of coupling ratio. Depending on the geometry and material characteristics, the presented model is able to simulate both hardening behavior (due to microstructure) and softening behavior (due to torsion/bending coupling) of the nanoscanners.

Ramifications of structural deformations on collapse loads of critically cracked pipe bends under in-plane bending and internal pressure

Energy Technology Data Exchange (ETDEWEB)

Sasidharan, Sumesh; Arunachalam, Veerappan; Subramaniam, Shanmugam [Dept. of Mechanical Engineering, National Institute of Technology, Tiruchirappalli (India)

2017-02-15

Finite-element analysis based on elastic-perfectly plastic material was conducted to examine the influence of structural deformations on collapse loads of circumferential through-wall critically cracked 90 .deg. pipe bends undergoing in-plane closing bending and internal pressure. The critical crack is defined for a through-wall circumferential crack at the extrados with a subtended angle below which there is no weakening effect on collapse moment of elbows subjected to in-plane closing bending. Elliptical and semioval cross sections were postulated at the bend regions and compared. Twice-elastic-slope method was utilized to obtain the collapse loads. Structural deformations, namely, ovality and thinning, were each varied from 0% to 20% in steps of 5% and the normalized internal pressure was varied from 0.2 to 0.6. Results indicate that elliptic cross sections were suitable for pipe ratios 5 and 10, whereas for pipe ratio 20, semioval cross sections gave satisfactory solutions. The effect of ovality on collapse loads is significant, although it cancelled out at a certain value of applied internal pressure. Thinning had a negligible effect on collapse loads of bends with crack geometries considered.

DEM study of granular discharge rate through a vertical pipe with a bend outlet in small absorber sphere system

Energy Technology Data Exchange (ETDEWEB)

Li, Tianjin, E-mail: tjli@tsinghua.edu.cn; Zhang, He; Liu, Malin; Huang, Zhiyong; Bo, Hanliang; Dong, Yujie

2017-04-01

Highlights: • The work concerns granular flow in a vertical pipe with a bend. • Discharge rate fluctuation in vertical pipe are mainly from velocity fluctuation. • Steady discharge rate decreases rapidly and saturates with μ{sub s} increasing. • Steady discharge rate W{sub s} still obey the 5/2 power law of pipe internal diameter. • A correlation developed for steady discharge rate for this new geometry. - Abstract: Absorber sphere pneumatic conveying is a special application of pneumatic conveying technique in the pebble bed High Temperature Gas-Cooled Reactor (HTGR or HTR). Granular discharge through a vertical pipe with a bend outlet is one of the control modes to determine solid mass flowrate which is an important parameter for the design of absorber sphere pneumatic conveying. Granular discharge rate through the vertical pipe with a bend outlet in the small absorber sphere system are investigated by discrete element method simulation. The effect of geometry parameters on discharge rate, the discharge rate fluctuation in the vertical pipe, and the effect of friction on steady discharge rate (W{sub s}) are analyzed and discussed. The phenomena of discharge rate fluctuation in the vertical pipe are observed, which are mainly resulted from the evolution of the average downward granular velocity. The steady discharge rate decreases rapidly with sliding friction coefficient increasing from 0.125 to 0.5, and gradually saturates with the friction coefficient further increasing from 0.5 to 1. It is interesting that the linear relation between W{sub s}{sup 2/5} and pipe internal diameter D with zero intercept are found for the vertical pipe discharge with a bend outlet, which is different from the orifice discharge through a hopper or silo with none-zero intercept. A correlation similar to Beverloo’s correlation is developed to predict the steady discharge rate through the vertical pipe with a bend outlet. These results are helpful for the design of sphere

Limit loads for pipe bends under combined pressure and in-plane bending based on finite element limit analysis

International Nuclear Information System (INIS)

Oh, Chang Sik; Kim, Yun Jae

2006-01-01

In the present paper, approximate plastic limit load solutions for pipe bends under combined internal pressure and bending are obtained from detailed three-dimensional (3-D) FE limit analyses based on elastic-perfectly plastic materials with the small geometry change option. The present FE results show that existing limit load solutions for pipe bends are lower bounds but can be very different from the present FE results in some cases, particularly for bending. Accordingly closed-form approximations are proposed for pipe bends under combined pressure and in-plane bending based on the present FE results. The proposed limit load solutions would be a basis of defective pipe bends and be useful to estimate non-linear fracture mechanics parameters based on the reference stress approach

Application of MCNPX 2.7.D for reactor core management at the research reactor BR2

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, Edgar

2011-01-01

The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK-CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full core model. The second method represents fully automatic whole core depletion and criticality calculations in the detailed 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared vs. the developed at the BR2 reactor department MCNPX & ORIGEN-S combined method. Testing of MCNPX 2.7.D on the criticality measurements at the BR2 reactor is presented. (author)

Study of Transport and Micro-structural properties of Magnesium Di-Boride Strand under react and bend mode and bend and react mode

International Nuclear Information System (INIS)

Kundu, Ananya; Kumar Das, Subrat; Bano, Anees; Pradhan, Subrata

2017-01-01

I-V characterization of commercial multi-filamentary Magnesium Di-Boride (MgB 2 ) wire of diameter 0.83 mm were studied in Cryocooler at self-field I-V characterization system under both react and bend mode and bend and react mode for a range of temperature 6 K - 25 K. This study is of practical technical relevance where the heat treatment of the superconducting wire makes the wire less flexible for winding in magnet and in other applications. In the present work the bending diameter was varied from 40 mm to 20 mm and for each case critical current (I c ) of the strand is measured for above range of temperature. A customized sample holder is fabricated and thermally anchored with the 2 nd cold stage of Cryocooler. It is observed from the measurement that the strand is more susceptible to degradation for react and bend cases. The transport measurement of the strand was accompanied by SEM analyses of bend samples. Also the tensile strength of the raw strands and the heat treated strands were carried out at room temperature in Universal Testing Machine (UTM) to have an estimate about the limiting winding tension value during magnet fabrication. (paper)

Fully integrated analysis of reactor kinetics, thermalhydraulics and the reactor control system in the MAPLE-X10 research reactor

International Nuclear Information System (INIS)

Shim, S.Y.; Carlson, P.A.; Baxter, D.K.

1992-01-01

A prototype research reactor, designated MAPLE-X10 (Multipurpose Applied Physics Lattice Experimental - X 10MW), is currently being built at AECL's Chalk River Laboratories. The CATHENA (Canadian Algorithm for Thermalhydraulic Network Analysis) two-fluid code was used in the safety analysis of the reactor to determine the adequacy of core cooling during postulated reactivity and loss-of-forced-flow transients. The system responses to a postulated transient are predicted including the feedback between reactor kinetics, thermalhydrauilcs and the reactor control systems. This paper describes the MAPLE-X10 reactor and the modelling methodology used. Sample simulations of postulated loss-of-heat-sink and loss-of-regulation transients are presented. (author)

Denitrification in the Mississippi River network controlled by flow through river bedforms

Science.gov (United States)

Gomez-Velez, Jesus D.; Harvey, Judson W.; Cardenas, M. Bayani; Kiel, Brian

2015-01-01

Increasing nitrogen concentrations in the world’s major rivers have led to over-fertilization of sensitive downstream waters1, 2, 3, 4. Flow through channel bed and bank sediments acts to remove riverine nitrogen through microbe-mediated denitrification reactions5, 6, 7, 8, 9, 10. However, little is understood about where in the channel network this biophysical process is most efficient, why certain channels are more effective nitrogen reactors, and how management practices can enhance the removal of nitrogen in regions where water circulates through sediment and mixes with groundwater - hyporheic zones8, 11, 12. Here we present numerical simulations of hyporheic flow and denitrification throughout the Mississippi River network using a hydrogeomorphic model. We find that vertical exchange with sediments beneath the riverbed in hyporheic zones, driven by submerged bedforms, has denitrification potential that far exceeds lateral hyporheic exchange with sediments alongside river channels, driven by river bars and meandering banks. We propose that geomorphic differences along river corridors can explain why denitrification efficiency varies between basins in the Mississippi River network. Our findings suggest that promoting the development of permeable bedforms at the streambed - and thus vertical hyporheic exchange - would be more effective at enhancing river denitrification in large river basins than promoting lateral exchange through induced channel meandering. 

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

International Nuclear Information System (INIS)

Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

2014-01-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

Energy Technology Data Exchange (ETDEWEB)

Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

2014-07-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

UNDERWATER ANALYSIS OF IRRADIATED REACTOR SLUGS FOR Co-60 AND OTHER RADIONUCLIDES AT THE SAVANNAH RIVER SITE

International Nuclear Information System (INIS)

CASELLA, VITO

2004-01-01

Co-60 was produced in the Savannah River Site (SRS) reactors in the 1970s, and the irradiated cobalt reactor slugs were stored in a reactor basin at SRS. Since the activity rates of these slugs were not accurately known, assaying was required. A sodium iodide gamma detector was placed above a specially designed air collimator assembly, so that the slug was eight to nine feet from the detector and was shielded by the basin water. Also, 18 curium sampler slugs, used to produce Cm-244 from Pu-239, were to be disposed of with the cobalt slugs. The curium slugs were also analyzed with a High Purity Germanium (HPGE) detector in an attempt to identify any additional radionuclides produced from the irradiation. Co-60 concentrations were determined for reactor disassembly basin cobalt slugs and the 18 curium sampler slugs. The total Co-60 activity of all of the assayed slugs in this work summed to 31,783 curies on 9/15/03. From the Co-60 concentrations of the curium sampler slugs, the irradiation flux was determined for the known irradiation time. The amounts of Pu-238,-239,-240,-241,-242; Am-241,-243; and Cm-242,-244 produced were then obtained based on the original amount of Pu-239 irradiated

Investigation of hydrodynamics on local scour by shape of single spur dike in river bend

International Nuclear Information System (INIS)

Masjedi, A; Foroushani, E P

2012-01-01

A series of experiments were conducted in which the the scour hole associated with model spur dike was measured in a 180 degree laboratory flume bend under clear-water overtopping flows. In this study, the local scour were conducted for three different shapes of oblong, rectangulat chamfered of straight spur dikes at the bend with various Froude number. The main goals of the experiments were to evaluate the effect of the three different shapes of straight spur dikes on the volume of scour and potential aquatic habitat and on minimizing erosion adjacent to the streambanks. The experiments showed that of the three different shapes of straight spur dikes tested, the least erosion of the around in the near bank region was associated with the spur dikes with oblong shape, while the greatest volume of the scour hole was associated with the rectangular shape. So it was observed that, as Froude number increases, the scour increases.

Core monitoring at the WNP-2 reactor

International Nuclear Information System (INIS)

Skeen, D.R.; Torres, R.H.; Burke, W.J.; Jenkins, I.; Jones, S.W.

1992-01-01

The WNP-2 reactor is a 3,323-MW(thermal) boiling water reactor (BWR) that is operated by the Washington Public Power Supply System. The WNP-2 reactor began commercial operation in 1984 and is currently in its eighth cycle. The core monitoring system used for the first cycle of operation was supplied by the reactor vendor. Cycles 2 through 6 were monitored with the POWERPLEX Core Monitoring Software System (CMSS) using the XTGBWR simulation code. In 1991, the supply system upgraded the core monitoring system by installing the POWERPLEX 2 CMSS prior to the seventh cycle of operation for WNP-2. The POWERPLEX 2 CMSS was developed by Siemens Power Corporation (SPC) and is based on SPC's advanced state-of-the-art reactor simulator code MICROBURN-B. The improvements in the POWERPLEX 2 system are possible as a result of advances in minicomputer hardware

Monitoring of irradiation effects on the pressure vessel steels of Calder, Chapelcross and Windscale Advanced Gas Cooled Reactor (WAGR) nuclear reactors

International Nuclear Information System (INIS)

Turner, F.

1980-01-01

Tensile, Charpy and bend specimens of plate, forging and weld metal are exposed in the lower and upper zones of the reactors to neutron fluxes covering the range experienced by the vessels. The test conditions are described, the results presented and discussed. (author)

Supplement to Final Environmental Statement related to construction and operation of Clinch River Breeder Reactor Plant, Docket No. 50-537

International Nuclear Information System (INIS)

1982-10-01

In February 1977, the Office of Nuclear Reactor Regulation issued a Final Environmental Statement (FES) (NUREG-0139) related to the construction and operation of the proposed Clinch River Breeder Reactor Plant (CRBRP). Since the FES was issued, additional data relative to the site and its environs have been collected, several modifications have been made to the CRBRP design, and its fuel cycle, and the timing of the plant construction and operation has been affected in accordance with deferments under the DOE Liquid Metal Fast Breeder Reactor (LMFBR) program. These changes are summarized and their environmental significance is assessed in this document. The reader should note that this document generally does not repeat the substantial amount of information in the FES which is still current; hence, the FES should be consulted for a comprehensive understanding of the staff's environmental review of the CRBRP project

Data management for the Clinch River Breeder Reactor Plant Project by use of document status and hold systems

International Nuclear Information System (INIS)

Hunt, C.S.; Beck, A.E.; Akhtar, M.S.

1982-01-01

This paper describes the development, framework, and scope of the Document Status System and the Document Hold System for the Clinch River Breeder Reactor Plant Project. It shows how data are generated at five locations and transmitted to a central computer for processing and storage. The resulting computerized data bank provides reports needed to perform day-to-day management and engineering planning. Those reports also partially satisfy the requirements of the Project's Quality Assurance Program

Mechanical design of a PERMCAT reactor module

Energy Technology Data Exchange (ETDEWEB)

Tosti, S. [Associazione ENEA Euratom sulla Fusione, C.R. ENEA Frascati, Via E. Fermi 45, Frascati, Roma I-00044 (Italy)], E-mail: tosti@frascati.enea.it; Bettinali, L. [Associazione ENEA Euratom sulla Fusione, C.R. ENEA Frascati, Via E. Fermi 45, Frascati, Roma I-00044 (Italy); Borgognoni, F. [Tesi Sas, Via Bolzano 28, Rome (Italy); Murdoch, D.K. [EFDA CSU, Boltzmannstr. 2, D-85748 Garching bei Munchen (Germany)

2007-02-15

The PERMCAT is a membrane reactor proposed for processing fusion reactor plasma exhaust gas: tritium removal is obtained by isotopic swamping operating in counter-current mode. In this work, a membrane reactor using a permeator tube of length about 500 mm produced via diffusion welding of Pd-Ag thin foils is described. An appropriate mechanical design of the membrane module has been developed in order to avoid any significant compressive and bending stresses on the very long and thin wall permeator tube: two expanded bellows have been applied to the Pd-Ag tube, so that it has been pre-tensioned before operating. The elongation of the metal permeator under hydrogenation has been theoretically estimated and experimentally verified for properly designing the membrane reactor.

Reactor enclosure. BRC meeting presentation

International Nuclear Information System (INIS)

Fisch, J.W.

1975-01-01

The latest status of key components of the Reactor Enclosure System of the Clinch River Breeder Reactor Plant is described. Areas where there have been notable design changes or significant design detail maturity in the six months since the last BRC presentation are highlighted. (auth)

Cadmium-emitter self-powered thermal neutron detector performance characterization & reactor power tracking capability experiments performed in ZED-2

Energy Technology Data Exchange (ETDEWEB)

LaFontaine, M.W., E-mail: physics@execulink.com [LaFontaine Consulting, Kitchener, Ontario (Canada); Zeller, M.B. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Nielsen, K. [Royal Military College of Canada, SLOWPOKE-2 Reactor, Kingston, Ontario (Canada)

2014-07-01

Cadmium-emitter self-powered thermal neutron flux detectors (SPDs), are typically used for flux monitoring and control applications in low temperature, test reactors such as the SLOWPOKE-2. A collaborative program between Atomic Energy of Canada, academia (Royal Military College of Canada (RMCC)) and industry (LaFontaine Consulting) was initiated to characterize the incore performance of a typical Cd-emitter SPD; and to obtain a definitive measure of the capability of the detector to track changes in reactor power in real time. Prior to starting the experiment proper, Chalk River Laboratories' ZED-2 was operated at low power (5 watts nominal) to verify the predicted moderator critical height. Test measurements were then performed with the vertical center of the SPD emitter positioned at the vertical mid-plane of the ZED-2 reactor core. Measurements were taken with the SPD located at lattice position L0 (near center), and repeated at lattice position P0 (in D{sub 2}O reflector). An ionization chamber (part of the ZED-2 control instrumentation) monitored reactor power at a position located on the south side of the outside wall of the reactor's calandria. These experiments facilitated measurement of the absolute thermal neutron sensitivity of the subject Cd-emitter SPD, and validated the power tracking capability of said SPD. Procedural details of the experiments, data, calculations and associated graphs, are presented and discussed. (author)

TA-2 Water Boiler Reactor Decommissioning Project

International Nuclear Information System (INIS)

Durbin, M.E.; Montoya, G.M.

1991-06-01

This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

Rupture prediction for induction bends under opening mode bending with emphasis on strain localization

International Nuclear Information System (INIS)

Mitsuya, Masaki; Sakanoue, Takashi

2015-01-01

This study focuses on the opening mode of induction bends; this mode represents the deformation outside a bend. Bending experiments on induction bends are shown and the manner of failure of these bends was investigated. Ruptures occur at the intrados of the bends, which undergo tensile stress, and accompany the local reduction of wall thickness, i.e., necking that indicates strain localization. By implementing finite element analysis (FEA), it was shown that the rupture is dominated not by the fracture criterion of material but by the initiation of strain localization that is a deformation characteristic of the material. These ruptures are due to the rapid increase of local strain after the initiation of strain localization and suddenly reach the fracture criterion. For the evaluation of the deformability of the bends, a method based on FEA that can predict the displacement at the rupture is proposed. We show that the yield surface shape and the true stress–strain relationship after uniform elongation have to be defined on the basis of the actual properties of the bend material. The von Mises yield criterion, which is commonly used in cases of elastic–plastic FEA, could not predict the rupture and overestimated the deformability. In contrast, a yield surface obtained by performing tensile tests on a biaxial specimen could predict the rupture. The prediction of the rupture was accomplished by an inverse calibration method that determined the true stress-strain relationship after uniform elongation. As an alternative to the inverse calibration, a simple extrapolation method of the true stress-strain relationship after uniform elongation which can predict the rupture is proposed. - Highlights: • A method based on FEA that can predict the displacement at the rupture is proposed. • The yield surface shape and the true stress–strain have to be defined precisely. • The von Mises yield criterion overestimated the deformability. • The ruptures are due to the

EBR-2 [Experimental Breeder Reactor-2] test programs

International Nuclear Information System (INIS)

Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

1990-01-01

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

Detrital zircon study along the Tsangpo River, SE Tibet

Science.gov (United States)

Liang, Y.; Chung, S.; Liu, D.; O'Reilly, S. Y.; Chu, M.; Ji, J.; Song, B.; Pearson, N. J.

2004-12-01

The interactions among tectonic uplift, river erosion and alluvial deposition are fundamental processes that shape the landscape of the Himalayan-Tibetan orogen since its creation from early Cenozoic time. To better understand these processes around the eastern Himalayan Syntaxis, we conducted a study by systematic sampling riverbank sediments along the Tsangpo River, SE Tibet. Detrital zircons separated from the sediments were subjected to U-Pb dating by the SHRIMP II at the Beijing SHRIMP Center and then in-situ measurements of Hf isotope ratios using LA-MC-ICPMS at GEMOC. These results, together with U-Pb ages and Hf isotope data that we recently obtained for the Transhimalayan plutonic and surrounding basement rocks, allow a more quantitative examination of the provenance or protosource areas for the river sediments. Consequently, the percentage inputs from these source areas can be estimated. Our study indicates that, before the Tsangpo River flows into the Namche Barwa Syntaxis of the eastern Himalayas where the River forms a 180° Big Bend gorge and crosscuts the Himalayan sequences, the Gangdese batholith that crops out just north of the River appear to be an overwhelming source accounting for ˜50 % of the bank sediments. The Tethyan Himalayan sequences south of the River are the second important source, with an input of ˜25 %. The proportion of sediment supply changes after the River enters the Big Bend gorge and turns to south: ˜25 % of detrital zircons are derived from the Greater Himalayas so that the input from the Tethyan Himalayas decreases (< 10 %) despite those from the Gangdese batholith remains high ( ˜40 %). Comparing with the sediment budget of the Brahmaputra River in the downstream based on literature Sr, Nd and Os isotope information, which suggests dominant ( ˜90-60 %) but subordinate ( ˜10-40 %) contributions by the (Greater and Lesser) Himalayan and Tibetan (including Tethyan Himalayan) rocks, respectively, the change is interpreted

Aerosol deposition in bends with turbulent flow

Energy Technology Data Exchange (ETDEWEB)

McFarland, A.R.; Gong, H.; Wente, W.B. [Texas A& M Univ., College Station, TX (United States)] [and others

1997-08-01

The losses of aerosol particles in bends were determined numerically for a broad range of design and operational conditions. Experimental data were used to check the validity of the numerical model, where the latter employs a commercially available computational fluid dynamics code for characterizing the fluid flow field and Lagrangian particle tracking technique for characterizing aerosol losses. Physical experiments have been conducted to examine the effect of curvature ratio and distortion of the cross section of bends. If it curvature ratio ({delta} = R/a) is greater than about 4, it has little effect on deposition, which is in contrast with the recommendation given in ANSI N13.1-1969 for a minimum curvature ratio of 10. Also, experimental results show that if the tube cross section is flattened by 25% or less, the flattening also has little effect on deposition. Results of numerical tests have been used to develop a correlation of aerosol penetration through a bend as a function of Stokes number (Stk), curvature ratio ({delta}) and the bend angle ({theta}). 17 refs., 10 figs., 2 tabs.

CRC DEPLETION CALCULATIONS FOR THE RODDED ASSEMBLIES IN BATCHES 1, 2, 3, AND 1X OF CRYSTAL RIVER UNIT 3

Energy Technology Data Exchange (ETDEWEB)

Kenneth D. Wright

1997-09-03

The purpose of this design analysis is to document the SAS2H depletion calculations of certain rodded fuel assemblies from batches 1, 2, 3, and 1X of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A rodded assembly is one that contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) for some period of time during its irradiation history. The objective of this analysis is to provide SAS2H calculated isotopic compositions of depleted fuel and depleted burnable poison for each fuel assembly to be used in subsequent CRC reactivity calculations containing the fuel assemblies.

CRC DEPLETION CALCULATIONS FOR THE RODDED ASSEMBLIES IN BATCHES 1, 2, 3, AND 1X OF CRYSTAL RIVER UNIT 3

International Nuclear Information System (INIS)

Wright, Kenneth D.

1997-01-01

The purpose of this design analysis is to document the SAS2H depletion calculations of certain rodded fuel assemblies from batches 1, 2, 3, and 1X of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A rodded assembly is one that contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) for some period of time during its irradiation history. The objective of this analysis is to provide SAS2H calculated isotopic compositions of depleted fuel and depleted burnable poison for each fuel assembly to be used in subsequent CRC reactivity calculations containing the fuel assemblies

Spent fuel strategy for the BR2 reactor

International Nuclear Information System (INIS)

Gubel, P.; Collard, G.

1998-01-01

The Belgian MTR reactor is fuelled with HEU UAl x elements and the fuel cycle was normally closed by reprocessing consecutively in Belgium (Eurochemic), France (Marcoule) and finally in the U.S.A. (Idaho Falls and Savannah River). When the acceptance of spent fuel by the U.S. was terminated, the facility was left with a huge backlog of used elements stored under water. After a few years, urgent and mandatory actions were required to maintain the BR2 facility operating. Later the accent was put on the evaluation of an optimum long term solution for the BR2 spent fuel during the projected 15 years life extension after the refurbishment executed between 1995 and 1997. The paper gives an overview of these successive actions taken during the last years as well as the handled various criteria for comparing and evaluating the available long-term alternatives. After commitment to reprocessing in existing facilities operated for aluminum fuels the focus of the BR2 fuel cycle strategy is now moving to the procurement of the necessary HEU fuel for securing the long-term operation of the facility. (author)

On the accuracy of analyses for in-plane bending of smooth pipe bends with end constraints

International Nuclear Information System (INIS)

Thomson, G.; Spence, J.

1985-01-01

The accuracy of theoretical analyses for in-plane bending of smooth pipebends with end constraints is discussed and investigated with a view to explaining and reducing the differences between the major works. An earlier theory of the authors is improved to give more accurate answers for bends with rigid flanges. Flanged bends are then examined in some detail, quantifying for the first time the important influence of the flange rigidity on the bend flexibility and stresses. A summary of some finite element analyses is presented from which it is clear that further work is desirable. (orig.)

Effect of bend separation distance on the mass transfer in back-to-back pipe bends arranged in a 180° configuration

Science.gov (United States)

Chen, X.; Le, T.; Ewing, D.; Ching, C. Y.

2016-12-01

The mass transfer to turbulent flow through back-to-back pipe bends arranged in a 180° configuration with different lengths of pipe between the bends was measured using a dissolving gypsum test section in water. The measurements were performed for bends with a radius of curvature of 1.5 times the pipe diameter ( D) at a Reynolds numbers of 70,000 and Schmidt number of 1280. The maximum mass transfer in the bends decreased from approximately 1.8 times the mass transfer in the upstream pipe when there was no separation distance between the bends to 1.7 times when there was a 1 D or 5 D length of pipe between the bends. The location of the maximum mass transfer was on the inner sidewall downstream of the second bend when there was no separation distance between the bends. This location changed to the inner wall at the beginning of the second bend when there was a 1 D long pipe between the bends, and to the inner sidewall at the end of the first bend when there was a 5 D long pipe between the bends.

Bend testing for miniature disks

International Nuclear Information System (INIS)

Huang, F.H.; Hamilton, M.L.; Wire, G.L.

1982-01-01

A bend test was developed to obtain ductility measurements on a large number of alloy variants being irradiated in the form of miniature disks. Experimental results were shown to be in agreement with a theoretical analysis of the bend configuration. Disk specimens fabricated from the unstrained grip ends of previously tested tensile specimens were used for calibration purposes; bend ductilities and tensile ductilities were in good agreement. The criterion for estimating ductility was judged acceptable for screening purposes

Transuranic radionuclides in the Columbia River: sources, inventories, and geochemical behavior

International Nuclear Information System (INIS)

Beasley, T.M.

1987-01-01

The sources, inventories, and geochemical behavior of transuranic and other long-lived radionuclides in the lower Columbia River are summarized. Inventories have been estimated from the measured activities of the different radionuclides in 50 cores raised in 1977 and 1978, while annual export of transuranic radionuclides was determined from monthly water collections in the estuary. Continental shelf inventories of Pu and Am isotopes have been estimated using excess 210 Pb inventories and the mean 210 Pb//sup 239,240/Pu inventory ratio of 100 +/- 19 observed in representative cores raised from the shelf. Despite the substantial past addition of radioactivity to the river from operation of the plutonium production reactors at Hanford, the amounts of reactor-derived radionuclides in river sediments are small relative to fallout-derived nuclides. Erosional processes have mobilized both fallout-derived /sup 239,240/Pu and 137 Cs from the landscape to the river, but the quantities involved represent <1% of their fallout inventories within the river's drainage basin. 36 references, 6 figures, 2 tables

Nuclear Regulatory Commission issuances

International Nuclear Information System (INIS)

1996-04-01

This report includes the issuances received during the April 1996 reporting period from the Commission, the Atomic Safety and Licensing Boards, the Administrative Law Judges, the Directors' Decisions, and the Decisions on Petitions for Rulemaking. Included are issuances pertaining to: (1) Yankee Nuclear Power Station, (2) Georgia Tech Research Reactor, (3) River Bend Station, (4) Millstone Unit 1, (5) Thermo-Lag fire barrier material, and (6) Louisiana Energy Services

Nuclear Regulatory Commission issuances

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-04-01

This report includes the issuances received during the April 1996 reporting period from the Commission, the Atomic Safety and Licensing Boards, the Administrative Law Judges, the Directors` Decisions, and the Decisions on Petitions for Rulemaking. Included are issuances pertaining to: (1) Yankee Nuclear Power Station, (2) Georgia Tech Research Reactor, (3) River Bend Station, (4) Millstone Unit 1, (5) Thermo-Lag fire barrier material, and (6) Louisiana Energy Services.

Determination of Columbia River flow times from Pasco, Washington using radioactive tracers introduced by the Hanford reactors

Science.gov (United States)

Nelson, Jack L.; Perkins, R.W.; Haushild, W.L.

1966-01-01

Radioactive tracers introduced into the Columbia River in cooling water from the Hanford reactors were used to measure flow times downstream from Pasco, Washington, as far as Astoria, Oregon. The use of two tracer methods was investigated. One method used the decay of a steady release of Na24 (15-hour half-life) to determine flow times to various downstream locations, and flow times were also determined from the time required for peak concentration of instantaneous releases of I131 (8-day half-life) to reach these locations. Flow times determined from the simultaneous use of the two methods agreed closely. The measured flow times for the 224 miles from Pasco to Vancouver, Washington, ranged from 14.6 to 3.6 days, respectively, for discharges of 108,000 and 630,000 ft3/sec at Vancouver, Washington. A graphic relation for estimating flow times at discharges other than those measured and for several locations between Pasco and Vancouver was prepared from the data of tests made at four river discharges. Some limited data are also presented on the characteristics of dispersion of I131 in the Columbia River.

Fabrication of topology optimized photonic crystal waveguide Z-bend displaying large bandwidth with very low bend loss

DEFF Research Database (Denmark)

Harpøth, Anders; Frandsen, Lars Hagedorn; Kristensen, Martin

2004-01-01

We have designed, simulated and fabricated a photonic crystal waveguide Z-bend, which displays a total bend loss of ~1dB per bend in a wavelength range of more than 200nm. The fabricated component performs in excellent agreement with 3D finite-difference time-domain calculations....

Channel modelling and performance analysis of V2I communication systems in blind bend scattering environments

KAUST Repository

Chelli, Ali; Hamdi, Rami; Alouini, Mohamed-Slim

2014-01-01

In this paper, we derive a new geometrical blind bend scattering model for vehicle-to- infrastructure (V2I) communications. The proposed model takes into account single-bounce and double- bounce scattering stemming from fixed scatterers located

The Spatial Distribution of Bed Sediment on Fluvial System: A Mini Review of the Aceh Meandering River

Directory of Open Access Journals (Sweden)

Muhammad Irham

2016-08-01

Full Text Available Dynamic interactions of hydrological and geomorphological processes in the fluvial system result in accumulated deposit on the bed because the capacity to carry sediment has been exceeded. The bed load of the Aceh fluvial system is primarily generated by mechanical weathering resulting in boulders, pebbles, and sand, which roll or bounce along the river bed forming temporary deposits as bars on the insides of meander bends, as a result of a loss of transport energy in the system. This dynamic controls the style and range of deposits in the Aceh River. This study focuses on the spatial distribution of bed-load transport of the Aceh River. Understanding the spatial distribution of deposits facilitates the reconstruction of the changes in controlling factors during accumulation of deposits. One of the methods can be done by sieve analysis of sediment, where the method illuminates the distribution of sediment changes associated with channel morphology under different flow regimes. Hence, the purpose of this mini review is to investigate how the sediment along the river meander spatially dispersed. The results demonstrate that channel deposits in the Aceh River are formed from four different type of materials: pebble deposited along upstream left bank; sand located on the upstream, downstream, and along meander belts; and silt and clay located along the cut bank of meander bends. Because of different depositional pattern, the distribution of the sediment along the river can be used as a surrogate to identify bank stability, as well as to predict critical geometry for meander bend initiation

BR2 Reactor: Irradiation of fuels

International Nuclear Information System (INIS)

Verwimp, A.

2005-01-01

Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

Distribution and geochemistry of selected trace elements in the Sacramento River near Keswick Reservoir

Science.gov (United States)

Antweiler, Ronald C.; Taylor, Howard E.; Alpers, Charles N.

2012-01-01

The effect of heavy metals from the Iron Mountain Mines (IMM) Superfund site on the upper Sacramento River is examined using data from water and bed sediment samples collected during 1996-97. Relative to surrounding waters, aluminum, cadmium, cobalt, copper, iron, lead, manganese, thallium, zinc and the rare-earth elements (REE) were all present in high concentrations in effluent from Spring Creek Reservoir (SCR), which enters into the Sacramento River in the Spring Creek Arm of Keswick Reservoir. SCR was constructed in part to regulate the flow of acidic, metal-rich waters draining the IMM Superfund site. Although virtually all of these metals exist in SCR in the dissolved form, upon entering Keswick Reservoir they at least partially converted via precipitation and/or adsorption to the particulate phase. In spite of this, few of the metals settled out; instead the vast majority was transported colloidally down the Sacramento River at least to Bend Bridge, 67. km from Keswick Dam.The geochemical influence of IMM on the upper Sacramento River was variable, chiefly dependent on the flow of Spring Creek. Although the average flow of the Sacramento River at Keswick Dam is 250m 3/s (cubic meters per second), even flows as low as 0.3m 3/s from Spring Creek were sufficient to account for more than 15% of the metals loading at Bend Bridge, and these proportions increased with increasing Spring Creek flow.The dissolved proportion of the total bioavailable load was dependent on the element but steadily decreased for all metals, from near 100% in Spring Creek to values (for some elements) of less than 1% at Bend Bridge; failure to account for the suspended sediment load in assessments of the effect of metals transport in the Sacramento River can result in estimates which are low by as much as a factor of 100. ?? 2012.

Untangling Trends and Drivers of Changing River Discharge Along Florida's Gulf Coast

Science.gov (United States)

Glodzik, K.; Kaplan, D. A.; Klarenberg, G.

2017-12-01

Along the relatively undeveloped Big Bend coastline of Florida, discharge in many rivers and springs is decreasing. The causes are unclear, though they likely include a combination of groundwater extraction for water supply, climate variability, and altered land use. Saltwater intrusion from altered freshwater influence and sea level rise is causing transformative ecosystem impacts along this flat coastline, including coastal forest die-off and oyster reef collapse. A key uncertainty for understanding river discharge change is predicting discharge from rainfall, since Florida's karstic bedrock stores large amounts of groundwater, which has a long residence time. This study uses Dynamic Factor Analysis (DFA), a multivariate data reduction technique for time series, to find common trends in flow and reveal hydrologic variables affecting flow in eight Big Bend rivers since 1965. The DFA uses annual river flows as response time series, and climate data (annual rainfall and evapotranspiration by watershed) and climatic indices (El Niño Southern Oscillation [ENSO] Index and North Atlantic Oscillation [NAO] Index) as candidate explanatory variables. Significant explanatory variables (one evapotranspiration and three rainfall time series) explained roughly 50% of discharge variation across rivers. Significant trends (representing unexplained variation) were shared among rivers, with geographical grouping of five northern rivers and three southern rivers, along with a strong downward trend affecting six out of eight systems. ENSO and NAO had no significant impact. Advancing knowledge of these dynamics is necessary for forecasting how altered rainfall and temperatures from climate change may impact flows. Improved forecasting is especially important given Florida's reliance on groundwater extraction to support its growing population.

Robotics at Savannah River

International Nuclear Information System (INIS)

Byrd, J.S.

1983-01-01

A Robotics Technology Group was organized at the Savannah River Laboratory in August 1982. Many potential applications have been identified that will improve personnel safety, reduce operating costs, and increase productivity using modern robotics and automation. Several active projects are under way to procure robots, to develop unique techniques and systems for the site's processes, and to install the systems in the actual work environments. The projects and development programs are involved in the following general application areas: (1) glove boxes and shielded cell facilities, (2) laboratory chemical processes, (3) fabrication processes for reactor fuel assemblies, (4) sampling processes for separation areas, (5) emergency response in reactor areas, (6) fuel handling in reactor areas, and (7) remote radiation monitoring systems. A Robotics Development Laboratory has been set up for experimental and development work and for demonstration of robotic systems

Geological Features Mapping Using PALSAR-2 Data in Kelantan River Basin, Peninsular Malaysia

Science.gov (United States)

Pour, A. B.; Hashim, M.

2016-09-01

In this study, the recently launched Phased Array type L-band Synthetic Aperture Radar-2 (PALSAR-2) onboard the Advanced Land Observing Satellite-2 (ALOS-2), remote sensing data were used to map geologic structural and topographical features in the Kelantan river basin for identification of high potential risk and susceptible zones for landslides and flooding areas. A ScanSAR and two fine mode dual polarization level 3.1 images cover Kelantan state were processed for comprehensive analysis of major geological structures and detailed characterizations of lineaments, drainage patterns and lithology at both regional and district scales. Red-Green-Blue (RGB) colour-composite was applied to different polarization channels of PALSAR-2 data to extract variety of geological information. Directional convolution filters were applied to the data for identifying linear features in particular directions and edge enhancement in the spatial domain. Results derived from ScanSAR image indicate that lineament occurrence at regional scale was mainly linked to the N-S trending of the Bentong-Raub Suture Zone (BRSZ) in the west and Lebir Fault Zone in the east of the Kelantan state. Combination of different polarization channels produced image maps contain important information related to water bodies, wetlands and lithological units for the Kelantan state using fine mode observation data. The N-S, NE-SW and NNE-SSW lineament trends were identified in the study area using directional filtering. Dendritic, sub-dendritic and rectangular drainage patterns were detected in the Kelantan river basin. The analysis of field investigations data indicate that many of flooded areas were associated with high potential risk zones for hydro-geological hazards such as wetlands, urban areas, floodplain scroll, meander bend, dendritic and sub-dendritic drainage patterns, which are located in flat topograghy regions. Numerous landslide points were located in rectangular drainage system that associated

GEOLOGICAL FEATURES MAPPING USING PALSAR-2 DATA IN KELANTAN RIVER BASIN, PENINSULAR MALAYSIA

Directory of Open Access Journals (Sweden)

A. B. Pour

2016-09-01

Full Text Available In this study, the recently launched Phased Array type L-band Synthetic Aperture Radar-2 (PALSAR-2 onboard the Advanced Land Observing Satellite-2 (ALOS-2, remote sensing data were used to map geologic structural and topographical features in the Kelantan river basin for identification of high potential risk and susceptible zones for landslides and flooding areas. A ScanSAR and two fine mode dual polarization level 3.1 images cover Kelantan state were processed for comprehensive analysis of major geological structures and detailed characterizations of lineaments, drainage patterns and lithology at both regional and district scales. Red-Green-Blue (RGB colour-composite was applied to different polarization channels of PALSAR-2 data to extract variety of geological information. Directional convolution filters were applied to the data for identifying linear features in particular directions and edge enhancement in the spatial domain. Results derived from ScanSAR image indicate that lineament occurrence at regional scale was mainly linked to the N-S trending of the Bentong-Raub Suture Zone (BRSZ in the west and Lebir Fault Zone in the east of the Kelantan state. Combination of different polarization channels produced image maps contain important information related to water bodies, wetlands and lithological units for the Kelantan state using fine mode observation data. The N-S, NE-SW and NNE-SSW lineament trends were identified in the study area using directional filtering. Dendritic, sub-dendritic and rectangular drainage patterns were detected in the Kelantan river basin. The analysis of field investigations data indicate that many of flooded areas were associated with high potential risk zones for hydro-geological hazards such as wetlands, urban areas, floodplain scroll, meander bend, dendritic and sub-dendritic drainage patterns, which are located in flat topograghy regions. Numerous landslide points were located in rectangular drainage system

Reactor core

International Nuclear Information System (INIS)

Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

1990-01-01

Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

Comments on nuclear reactor safety in Ontario

International Nuclear Information System (INIS)

1987-08-01

The Chalk River Technicians and Technologists Union representing 500 technical employees at the Chalk River Nuclear Laboratories of AECL submit comments on nuclear reactor safety to the Ontario Nuclear Safety Review. Issues identified by the Review Commissioner are addressed from the perspective of both a labour organization and experience in the nuclear R and D field. In general, Local 1568 believes Ontario's CANDU nuclear reactors are not only safe but also essential to the continued economic prosperity of the province

Ageing management of the BR2 research reactor

International Nuclear Information System (INIS)

Verpoortem, J. R.; Van Dyck, S.

2014-01-01

At the Belgian nuclear research centre (SCK.CEN) several test reactors are operated. Among these, Belgian Reactor 2 (BR2) is the largest Material Test Reactor (MTR). This water-cooled, beryllium moderated reactor with a maximum thermal power of 100 MW became operational in 1962. Except for two major refurbishment campaigns of one year each, this reactor has been operated continuously over the past 50 years, with a frequency of 5-12 cycles per year. At present, BR2 is used for different research activities, the production of medical isotopes, the production of n-doped silicon and various training and education activities. (Author)

Seismic evaluation of safety systems at the Savannah River reactors

International Nuclear Information System (INIS)

Hardy, G.S.; Johnson, J.J.; Eder, S.J.; Monahon, T.M.; Ketcham, D.R.

1989-01-01

A thorough review of all safety related systems in commercial nuclear power plants was prompted by the accident at the Three Mile Island Nuclear Power Plant. As a consequence of this review, the Nuclear Regulatory Commission (NRC) focused its attention on the environmental and seismic qualification of the industry's electrical and mechanical equipment. In 1980, the NRC issued Unresolved Safety Issue (USI) A-46 to verify the seismic adequacy of the equipment required to safely shut down a plant and maintain a stable condition for 72 hours. After extensive research by the NRC, it became apparent that traditional analysis and testing methods would not be a feasible mechanism to address this USI A-46 issue. The costs associated with utilizing the standard analytical and testing qualification approaches were exorbitant and could not be justified. In addition, the only equipment available to be shake table testing which is similar to the item being qualified is typically the nuclear plant component itself. After 8 years of studies and data collection, the NRC issued its ''Generic Safety Evaluation Report'' approving an alternate seismic qualification approach based on the use of seismic experience data. This experience-based seismic assessment approach will be the basis for evaluating each of the 70 pre-1972 commercial nuclear power units in the United States and for an undetermined number of nuclear plants located in foreign countries. This same cost-effective developed for the commercial nuclear power industry is currently being applied to the Savannah River Production Reactors to address similar seismic adequacy issues. This paper documents the results of the Savannah River Plant seismic evaluating program. This effort marks the first complete (non-trial) application of this state-of-the-art USI A-46 resolution methodology

Seismic evaluation of safety systems at the Savannah River reactors

International Nuclear Information System (INIS)

Hardy, G.S.; Johnson, J.J.; Eder, S.J.; Monahon, T.; Ketcham, D.

1989-01-01

A thorough review of all safety related systems in commercial nuclear power plants was prompted by the accident at the Three Mile Island Nuclear Power Plant. As a consequence of this review, the Nuclear Regulatory Commission (NRC) focused its attention on the environmental and seismic qualification of the industry's electrical and mechanical equipment. In 1980, the NRC issued Unresolved Safety Issue (USI) A-46 to verify the seismic adequacy of the equipment required to safely shut down a plant and maintain a stable condition for 72 hours. After extensive research by the NRC, it became apparent that traditional analysis and testing methods would not be a feasible mechanism to address this USI A-46 issue. The costs associated with utilizing the standard analytical and testing qualification approaches were exorbitant and could not be justified. In addition, the only equipment available to be shake table tested which is similar to the item being qualified is typically the nuclear plant component itself. After 8 years of studies and data collection, the NRC issued its Generic Safety Evaluation Report approving an alternate seismic qualification approach based on the use of seismic experience data. This experience-based seismic assessment approach will be the basis for evaluating each of the 70 pre-1972 commercial nuclear power units in the US and for an undetermined number of nuclear plants located in foreign countries. This same cost-effective approach developed for the commercial nuclear power industry is currently being applied to the Savannah River Production Reactors to address similar seismic adequacy issues. This paper documents the results of the Savannah River Plant seismic evaluation program. This effort marks the first complete (non-trial) application of this state-of-the-art USI A-46 resolution methodology

Does the Wuergassen reactor contribute to forest disease in the Solling?

International Nuclear Information System (INIS)

Streletzki, H.W.

1987-01-01

On the basis of a survey of forest diseases conducted in 1983 around the Wuergassen reactor, Reichelt in 1985 arrived at the conclusion that his hypothesis of 'the high level of forest disease observed in the area of the bend of the river Weser being mainly caused by the reactor' must in all probability be accurate. Thanks to the vast amount of data provided by the aerial photography evaluation, that hypothesis could be verified in 164 sampling points. The results did not confirm any significant influence of the Wuergassen reactor on the indices of harm in the area of the main wind direction. Instead it was found in the area of investigation II, projecting some 18 kilometres into the Solling mountain, that damage decreases from the marginal part to the areas further away to the east. Within that range of distance, the age of stands and their altitude could be clearly identified as factors having a significant influence on the index of harm. Consequently, although the study confirmed the decrease of harm from west to east observed by Reichelt, the highest incidence of damage was found on the eastern fringe and not only in the main wind direction of the reactor. This result led to the hypothesis that large-scale factors of harm act on the western fringe of the Solling, not the reactor. To verify that hypothesis, the north-western part of the Solling (area of investigation III) was included in the evaluation to the depth of, equally, 18 kilometres. The evaluation confirmed the assumptions formulated in the hypothesis. An influence of the reactor on forest disease in the south-western part of the Solling is to be excluded on the basis of these investigation results. Instead, factors of harm acting on a large scale, for instance pollutant burdens carried into the region from remote sources, must be considered as an essential cause of the increased incidence of harm in the entire western fringe of the Solling forest. (orig.) [de

Savannah River Site reactor hardware design modification study

International Nuclear Information System (INIS)

Fisher, J.E.

1990-03-01

A study was undertaken to assess the merits of proposed design modifications to the SRS reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. For the elevated piping design, system recovery was predicted for breaks in the plenum inlet or pump suction piping; response to the pump discharge break location did not show improvement compared to the present system configuration. The rotovalve closure design improved system response to plenum inlet or pump discharge breaks; recovery was not predicted for pump suction breaks. The pump suction valve closure design demonstrated system recovery for all break locations downstream of the valve. A combination of features is recommended to ensure liquid inventory recovery for all break locations. The elevated piping design performance during pump discharge breaks would be improved with addition of a dc pump trip in the affected loop. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit from the clamshell enclosing that section of piping. 12 refs., 10 figs., 2 tabs

Machinery Vibration Monitoring Program at the Savannah River Site

International Nuclear Information System (INIS)

Potvin, M.M.

1990-01-01

The Reactor Maintenance's Machinery Vibration Monitoring Program (MVMP) plays an essential role in ensuring the safe operation of the three Production Reactors at the Westinghouse Savannah River Company (WRSC) Savannah River Site (SRS). This program has increased machinery availability and reduced maintenance cost by the early detection and determination of machinery problems. This paper presents the Reactor Maintenance's Machinery Vibration Monitoring Program, which has been documented based on Electric Power Research Institute's (EPRI) NP-5311, Utility Machinery Monitoring Guide, and some examples of the successes that it has enjoyed

The investigation of multi-channel splitters and big-bend waveguides based on 2D sunflower-typed photonic crystals

Science.gov (United States)

Liu, Wei; Sun, XiaoHong; Fan, QingBin; Wang, Shuai; Qi, YongLe

2016-12-01

Different kinds of multi-channel splitters and big-bend waveguides have been designed and investigated by using sunflower-typed photonic crystals. By comparing the transmission spectra of two kinds of 4-channels beam splitters, we find that "C" type splitter has a relative uniform splitting ratio for different channels in a certain wavelength range. Furthermore three types of waveguides with different bending degrees have been investigated. Except for a little loss in the short wavelength with the increase of the bending degrees, they have almost the same transmission spectra structures. The result can be extended to big-bend waveguides with arbitrary bending degrees. This research is valuable for developing new-typed integrated optical communication devices.

Development and testing of the EDF-2 reactor fuel element

International Nuclear Information System (INIS)

Delpeyroux, P.

1964-01-01

This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the β phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors) [fr

Occipital bending in depression.

Science.gov (United States)

Maller, Jerome J; Thomson, Richard H S; Rosenfeld, Jeffrey V; Anderson, Rodney; Daskalakis, Zafiris J; Fitzgerald, Paul B

2014-06-01

There are reports of differences in occipital lobe asymmetry within psychiatric populations when compared with healthy control subjects. Anecdotal evidence and enlarged lateral ventricles suggests that there may also be a different pattern of curvature whereby one occipital lobe wraps around the other, termed 'occipital bending'. We investigated the prevalence of occipital bending in 51 patients with major depressive disorder (males mean age = 41.96 ± 14.00 years, females mean age = 40.71 ± 12.41 years) and 48 age- and sex-matched healthy control subjects (males mean age = 40.29 ± 10.23 years, females mean age = 42.47 ± 14.25 years) and found the prevalence to be three times higher among patients with major depressive disorder (18/51, 35.3%) when compared with control subjects (6/48, 12.5%). The results suggest that occipital bending is more common among patients with major depressive disorder than healthy subjects, and that occipital asymmetry and occipital bending are separate phenomena. Incomplete neural pruning may lead to the cranial space available for brain growth being restricted, or ventricular enlargement may exacerbate the natural occipital curvature patterns, subsequently causing the brain to become squashed and forced to 'wrap' around the other occipital lobe. Although the clinical implications of these results are unclear, they provide an impetus for further research into the relevance of occipital bending in major depression disorder. © The Author (2014). Published by Oxford University Press on behalf of the Guarantors of Brain. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Enhanced fluorescence detection of dansyl derivatives of phenolic compounds using a postcolumn photochemical reactor and application to chlorophenols in river water

Energy Technology Data Exchange (ETDEWEB)

de Ruiter, C.; Bohle, J.F.; de Jong, G.J.; Brinkman, U.A.T.; Frei, R.W.

1988-04-01

Photochemical decomposition by ultraviolet (UV) irradiation of dansyl derivatives of phenolic compounds in methanol-water mixtures leads to the formation of highly fluorescent dansyl-OH and dansyl-OCH/sub 3/. With substituted phenols as model compounds, it is demonstrated that inductive effects, caused by the substituents, play a major role in the gain in fluorescence signal (up to 8000-fold) that is obtained after postcolumn UV irradiation of the dansyl derivative, compared to that of the nonirradiated derivative. The optimal irradiation time for the dansyl derivatives is about 5.5 s. All monosubstituted phenolic dansyl derivatives now have a comparable limit of detection of approximately 200 pg (S/N = 3). The calibration curve of dansylated pentachlorophenol, using the postcolumn photochemical reactor under optimal conditions, is linear over at least 3 orders of magnitude with a correlation coefficient of 0.9999 (n = 9). Application of the system to the liquid chromatographic determination of highly chlorinated phenols in river water is presented. The repeatability of the system for a river water sample, spiked with 1 ppb pentachlorophenol, is 2.4% relative standard deviation (n = 5).

Reactor FaceMap Tool: A modern graphics tool for displaying reactor data

International Nuclear Information System (INIS)

Roberts, J.C.

1991-01-01

A prominent graphical user interface in reactor physics applications at the Savannah River Site is the reactor facemap display. This is a two dimensional view of a cross section of a reactor. In the past each application which needed a facemap implemented its own version. Thus, none of the code was reused, the facemap implementation was hardware dependent and the user interface was different for each facemap. The Reactor FaceMap Tool was built to solve these problems. Through the use of modern computing technologies such as X Windows, object-oriented programming and client/server technology the Reactor FaceMap Tool has the flexibility to work in many diverse applications and the portability to run on numerous types of hardware

Shipment of Taiwanese research reactor spent nuclear fuel (Phase 2): Environmental assessment

International Nuclear Information System (INIS)

1988-06-01

The proposed action is to transport approximately 1100 spent fuel rods from a foreign research reactor in Taiwan by sea to Hampton Roads, Virginia, and then overland by truck to the receiving basin for offsite fuels at the Savannah River Plant (SRP) for reprocessing to recover uranium and plutonium. The analysis of the impacts of the proposed action have been evaluated and shown to have negligible impact on the local environments. The calculations have been completed using the RADTRAN III code. PWR spent fuel was analyzed as a benchmark to link the calculations in this analysis to those in earlier environmental documentation. Cumulative total, maximum annual, and per shipment risks were calculated. The results indicate that the PWR spent fuel shipment risks are somewhat lower than those previously estimated. The cumulative and maximum annual normal, or incident-free, risks associated with the shipment of Taiwanese research reactor spent fuel is a factor of 10 lower than that for PWR fuel, and the cumulative and maximum annual accident radiological risks are a factor of about 2.2 lower than that for PWR spent fuel. As a result, the port risks are about a factor of 10 larger than the risk of overland transport. All of the risks calculated are small. The PWR risk values are similar to those judged by the NRC to be small enough not to warrant increased stringency in regulations. The Taiwanese research reactor spent fuel shipment risk values are smaller yet. 51 refs., 22 tabs

Design and Construction of the Plat Bending Machine

International Nuclear Information System (INIS)

Edy Sumarno; Abdul Hafid; Ismu H; Joko P W; Bambang Heru

2003-01-01

The plat-bending machine has been fabricated. The type is manual. That machine was made by plate, cylinder and U plat material. The machine has dimensions 110 mm in height, 650 mm in width, and 1200 mm in height. The capability of this machine is bending the plat with 2 mm in thickness and 1000 mm in width. This machine has the advantage to operate without electrical supply and easy to operate. (author)

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Analysis of coolability of the control rods of a Savannah River Site production reactor with loss of normal forced convection cooling

International Nuclear Information System (INIS)

Easterling, T.C.; Hightower, N.T.; Smith, D.C.; Amos, C.N.

1992-01-01

An analytical study of the coolability of the control rods in the Savannah River Site (SRS) K-Production Reactor under conditions of loss of normal forced convection cooling has been performed. The study was performed as part of the overall safety analysis of the reactor supporting its restart. The analysis addresses the buoyancy-driven flow over the control rods that occurs when forced cooling is lost, and the limit of critical heat flux that sets the acceptance criteria for the study. The objective of the study is to demonstrate that the control rods will remain cooled at powers representative of those anticipated for restart of the reactor. The study accomplishes this objective with a very tractable simplified analysis for the modest restart power. In addition, a best-estimate calculation is performed, and the results are compared to results from sub-scale scoping experiments. 5 refs

Cumulative damage fatigue tests on nuclear reactor Zircaloy-2 fuel tubes at room temperature and 3000C

International Nuclear Information System (INIS)

Pandarinathan, P.R.; Vasudevan, P.

1980-01-01

Cumulative damage fatigue tests were conducted on the Zircaloy-2 fuel tubes at room temperature and 300 0 C on the modified Moore type, four-point-loaded, deflection-controlled, rotating bending fatigue testing machine. The cumulative cycle ratio at fracture for the Zircaloy-2 fuel tubes was found to depend on the sequence of loading, stress history, number of cycles of application of the pre-stress and the test temperature. A Hi-Lo type fatigue loading was found to be very much damaging at room temperature and this feature was not observed in the tests at 300 0 C. Results indicate significant differences in damage interaction and damage propagation under cumulative damage tests at room temperature and at 300 0 C. Block-loading fatigue tests are suggested as the best method to determine the life-time of Zircaloy-2 fuel tubes under random fatigue loading during their service in the reactor. (orig.)

Experimental Determination of Bending Resonances of Millimeter Size PVF2 Cantilevers

Directory of Open Access Journals (Sweden)

David F. Thompson

2003-07-01

Full Text Available The polymer piezoelectric polvinylidene fluoride has found widespread use in sensors and actuators. The bending mode of piezoelectricity offers very high sensitivities and low mechanical input impedance, but has not been studied in as much detail for sensor applications. We report the dynamic electromechanical properties of millimeter size cantilevers made from electroded films of PVF2. All devices tested had a single polymer layer. Several resonances are found below 1 kHz and the experimentally observed resonance frequency dependence on cantilever thickness and length are seen to agree well with published models which take the properties of the electrodes into account. It is found that bending resonances are also modulated by the width of the cantilever. Therefore, though the length and thickness control the resonance frequency most strongly, the actual realized value can be fine-tuned by changing cantilever width and the electrode material and its thickness. Further, all resonances display high piezoelectric coupling coefficients (keff, ranging between 0.2 - 0.35. The data presented here will be extremely useful in the design of sensors and actuators for a number of applications, since the combination of millimeter size scales and high piezoelectric sensitivities in the low audio range can be realized with this marriage of polymeric materials and cantilever geometries. Such an array of sensors can be used in cochlear implant applications, and when integrated with a resonance interrogation circuit can be used for the detection of low frequency vibrations of large structures. If appropriate mass/elasticity sensitive layers are coated on the electrodes, such a sensor can be used for the detection of a wide range of chemicals and biochemicals.

Selective degradation of ibuprofen and clofibric acid in two model river biofilm systems.

Science.gov (United States)

Winkler, M; Lawrence, J R; Neu, T R

2001-09-01

A field survey indicated that the Elbe and Saale Rivers were contaminated with both clofibric acid and ibuprofen. In Elbe River water we could detect the metabolite hydroxy-ibuprofen. Analyses of the city of Saskatoon sewage effluent discharged to the South Saskatchewan river detected clofibric acid but neither ibuprofen nor any metabolite. Laboratory studies indicated that the pharmaceutical ibuprofen was readily degraded in a river biofilm reactor. Two metabolites were detected and identified as hydroxy- and carboxy-ibuprofen. Both metabolites were observed to degrade in the biofilm reactors. However, in human metabolism the metabolite carboxy-ibuprofen appears and degrades second whereas the opposite occurs in biofilm systems. In biofilms the pharmacologically inactive stereoisomere of ibuprofen is degraded predominantly. In contrast, clofibric acid was not biologically degraded during the experimental period of 21 days. Similar results were obtained using biofilms developed using waters from either the South Saskatchewan or Elbe River. In a sterile reactor no losses of ibuprofen were observed. These results suggested that abiotic losses and adsorption played only a minimal role in the fate of the pharmaceuticals in the river biofilm reactors.

Safety evaluation report related to the construction of the Clinch River Breeder Reactor Plant. Docket No. 50-537. Suppl. 1

International Nuclear Information System (INIS)

1983-05-01

Since the preparation of the Safety Evaluation Report the Advisory Committee on Reactor Safeguards considered the Clinch River construction permit license application at its 276th meeting and subsequently issued a favorable report, dated April 19, 1983 to the Commission (See Appendix I of this report). Additional documents associated with the application have been reviewed and a number of meetings have been held with the applicants. These events and documents are identified in Appendix E to this supplement. This supplement, SSER-1, to the Safety Evaluation Report, provides an evaluation of additional information received from the applicants since preparation of the SER regarding previously identified outstanding review items, and our response to the comments made by the Advisory Committee on Reactor Safeguards in its report

Use of Infrasound for evaluating potentially hazardous conditions for barge transit on the Mississippi River at Vicksburg, Mississippi

Science.gov (United States)

McKenna, M. H.; Simpson, C. P.; Jordan, A. M.

2017-12-01

Navigating the Mississippi River in Vicksburg, MS is known to be difficult for barge traffic in even the best of conditions due to the river's sharp bend 2 km north of the Highway 80 Bridge. When river levels rise, the level of difficulty in piloting barges under the bridge rises. Ongoing studies by the U.S. Army Engineer Research and Development Center (ERDC) are investigating infrasound as a means to correlate the low frequency acoustics generated by the river with the presence of hazardous conditions observed during flood stage, i.e., rough waters and high currents, which may lead to barge-bridge impacts. The Denied Area Monitoring and Exploitation of Structures (DAMES) Array at the ERDC Vicksburg, MS campus is a persistent seismic-acoustic array used for structural monitoring and explosive event detection. The DAMES Array is located 4.3 km from the Mississippi River/Highway 80 Bridge junction and recorded impulsive sub-audible acoustic signals, similar to an explosive event, from barge-bridge collisions that occurred between 2011 and 2017. This study focuses on five collisions that occurred during January 2016, which resulted in closing the river for barge transit and the Highway 80 Bridge for rail transit for multiple days until safety inspections were completed. The Highway 80 Bridge in Vicksburg, MS is the only freight-crossing over the Mississippi River between Baton Rouge, LA and Memphis, TN, meaning delays from these closings have significant impacts on all transit of goods throughout the Southeastern United States. River basin data and regional meteorological data have been analyzed to find correlations between the river conditions in January 2016, and recorded infrasound data with the aim of determining the likelihood that hazardous conditions are present on the river. Frequency-wavenumber analysis was used to identify the transient signals associated with the barge-bridge impacts and calculate the backazimuth to their source. Then, with the use of

Probabilistic evaluation of main coolant pipe break indirectly induced by earthquakes Savannah River Project L and P Reactors

International Nuclear Information System (INIS)

Short, S.A.; Wesley, D.A.; Awadalla, N.G.; Kennedy, R.P.

1989-01-01

A probabilistic evaluation of seismically-induced indirect pipe break for the Savannah River Project (SRP) L- and P-Reactor main coolant (process water) piping has been conducted. Seismically-induced indirect pipe break can result primarily from: (1) failure of the anchorage of one or more of the components to which the pipe is anchored; or (2) failure of the pipe due to collapse of the structure. the potential for both types of seismically-induced indirect failures was identified during a seismic walkdown of the main coolant piping. This work involved: (1) identifying components or structures whose failure could result in pipe failure; (2) developing seismic capacities or fragilities of these components; (3) combining component fragilities to develop plant damage state fragilities; and (4) convolving the plant seismic fragilities with a probabilistic seismic hazard estimate for the site in order to obtain estimates of seismic risk in terms of annual probability of seismic-induced indirect pipe break

Nuclear engineering R ampersand D at the Savannah River Site

International Nuclear Information System (INIS)

Strosnider, D.R.; Ferrara, W.R.

1991-01-01

The Westinghouse Savannah River Company (WSRC) is the prime operating contractor for the US Department of Energy at the Savannah River Site (SRS), located near Aiken, South Carolina. One division of WSRC, the Savannah River Laboratory (SRL), has the primary responsibility for research and development, which includes supporting the safe and efficient operation of the SRS production reactors. Several Sections of SRL, as well as other organization in WSRC, pursue R ampersand D and oversight activities related to nuclear engineering. The Sections listed below are described in more detail in this document: (SRL) nuclear reactor technology and scientific computations department; (SRL) safety analysis and risk management department; (WSRC) new production reactor program; and (WSRC) environment, safety, health, and quality assurance division

N Reactor thermal plume characterization during Pu-only mode of operation

Energy Technology Data Exchange (ETDEWEB)

Ecker, R.M.; Thompson, F.L.; Whelan, G.

1983-04-01

Pacific Northwest Laboratories (PNL) performed field and modeling studies -from March 1982 through June 1983 to characterize the thermal plume from the N Reactor heated water outfall while the N Reactor operated in the Pu-only mode. Part 1 of this report deals with the field studies conducted to characterize the N Reactor thermal plume while in the Pu-only mode of operation. It includes a description of the study area, a description of field tasks and procedures, and data collection results and discussion. Part 2 describes the computer simulation of the thermal plume under different flow conditions and the calibration of the model used. It includes a description of the computer model and the assumptions on which it is based, a presentation of the input data used in this application, and a discussion of modeling results. Because the field studies were restricted by the NPOES permit variance to the spring months when high Columbia River flows prevail the mathematical modeling of the N Reactor thermal plume while the reactor operates in the Pu-only mode is instrumental in characterizing the plume during low Columbia River flows.

Post-Columbia River Basalt Group stratigraphy and map compilation of the Columbia Plateau, Oregon

International Nuclear Information System (INIS)

Farooqui, S.M.; Bunker, R.C.; Thoms, R.E.; Clayton, D.C.; Bela, J.L.

1981-01-01

This report presents the results of reconnaissance mapping of sedimentary deposits and volcanic rocks overlying the Columbia River Basalt. The project area covers parts of the Dalles, Pendleton, Grangeville, Baker, Canyon City, and Bend. The mapping was done to provide stratigraphic data on the sedimentary deposits and volcanic rocks overlying the Columbia River Basalt Group. 160 refs., 16 figs., 1 tab

Environmental characterization to assess potential impacts of thermal discharge to the Columbia River

International Nuclear Information System (INIS)

Neitzel, D.A.; Dauble, D.D.; Page, T.L.; Greager, E.M.

1990-01-01

Laboratory and field studies were conducted to assess the potential impact of the N-Reactor thermal plume on fish from the Hanford Reach of the Columbia River. Discharge water temperatures were measured over a range of river flows and reactor operating conditions. Data were mathematically modeled to define spatial and thermal characteristics of the plume. Four species of Columbia River fish were exposed to thermal conditions expected in the plume. Exposed fish were subjected to predators and disease organisms to test for secondary effects from thermal stress. Spatial and temporal distribution of anadromous fish in the river near N-Reactor were also evaluated to define location relative to the plume. Potential thermal exposures were insufficient to kill or injure fish during operation of N-Reactor. These studies demonstrate that characterization of hydrological conditions and thermal tolerance can adequately assess potential impacts of a thermal discharge to fish

The novel ethylene-responsive factor CsERF025 affects the development of fruit bending in cucumber.

Science.gov (United States)

Wang, Chunhua; Xin, Ming; Zhou, Xiuyan; Liu, Chunhong; Li, Shengnan; Liu, Dong; Xu, Yuan; Qin, Zhiwei

2017-11-01

Overexpression of CsERF025 induces fruit bending by promoting the production of ethylene. Cucumber fruit bending critically affects cucumber quality, but the mechanism that causes fruit bending remains unclear. To better understand this mechanism, we performed transcriptome analyses on tissues from the convex (C1) and concave (C2) sides of bending and straight (S) fruit at 2 days post anthesis (DPA). We identified a total of 281 differentially expressed genes (DEGs) from both the convex and concave sides of bent fruit that showed significantly different expression profiles relative to straight fruits. Of these 281 DEGs, 196 were up-regulated (C1/S_C2/S) and 85 were down-regulated (C1/S_C2/S). Among the 196 up-regulated DEGs, the transcriptional levels of genes related to ethylene biosynthesis and signaling pathways were significantly higher in bending fruit compared with straight fruit. CsERF025 showed the largest difference in expression between bending and straight fruit. CsERF025 is an AP2/ERF gene encoding a protein that localizes to the nucleus. Overexpression of this gene increased the bending rate of cucumber fruits and increased the angle of bending. CsERF025 increased both the expression of ethylene biosynthesis-related genes and the production of ethylene. The application of exogenous 1-aminocyclopropane-l-carboxylic acid (ACC) to straight fruits from control plants promoted fruit bending. Thus, CsERF025 enhances the production of ethylene and thereby promotes fruit bending in cucumber.

Reliability of non-heated tube bends of boilers

International Nuclear Information System (INIS)

Bugaj, N.V.; Akhremenko, V.L.; Zamotaev, V.S.

1984-01-01

Bend failures are described for non-heated boiler tubes of 12Kh1MF and 20 steels. Methods of reliability evaluations are presented which permit revealing and replacing the bends with inadequate resources. Influences of operation conditions on bend durability is shown as well as the factors which are dominating at bend failures

Bending sound in graphene: Origin and manifestation

Energy Technology Data Exchange (ETDEWEB)

Adamyan, V.M., E-mail: vadamyan@onu.edu.ua [Department of Theoretical Physics, Odessa I.I. Mechnikov National University, 2 Dvoryanska St., Odessa 65026 (Ukraine); Bondarev, V.N., E-mail: bondvic@onu.edu.ua [Department of Theoretical Physics, Odessa I.I. Mechnikov National University, 2 Dvoryanska St., Odessa 65026 (Ukraine); Zavalniuk, V.V., E-mail: vzavalnyuk@onu.edu.ua [Department of Theoretical Physics, Odessa I.I. Mechnikov National University, 2 Dvoryanska St., Odessa 65026 (Ukraine); Department of Fundamental Sciences, Odessa Military Academy, 10 Fontanska Road, Odessa 65009 (Ukraine)

2016-11-11

Highlights: • The origin of sound-like dispersion of graphene bending mode is disclosed. • The speed of graphene bending sound is determined. • The renormalized graphene bending rigidity is derived. • The intrinsic corrugations of graphene are estimated. - Abstract: It is proved that the acoustic-type dispersion of bending mode in graphene is generated by the fluctuation interaction between in-plane and out-of-plane terms in the free energy arising with account of non-linear components in the graphene strain tensor. In doing so we use an original adiabatic approximation based on the alleged (confirmed a posteriori) significant difference of sound speeds for in-plane and bending modes. The explicit expression for the bending sound speed depending only on the graphene mass density, in-plane elastic constants and temperature is deduced as well as the characteristics of the microscopic corrugations of graphene. The obtained results are in good quantitative agreement with the data of real experiments and computer simulations.

Bending sound in graphene: Origin and manifestation

International Nuclear Information System (INIS)

Adamyan, V.M.; Bondarev, V.N.; Zavalniuk, V.V.

2016-01-01

Highlights: • The origin of sound-like dispersion of graphene bending mode is disclosed. • The speed of graphene bending sound is determined. • The renormalized graphene bending rigidity is derived. • The intrinsic corrugations of graphene are estimated. - Abstract: It is proved that the acoustic-type dispersion of bending mode in graphene is generated by the fluctuation interaction between in-plane and out-of-plane terms in the free energy arising with account of non-linear components in the graphene strain tensor. In doing so we use an original adiabatic approximation based on the alleged (confirmed a posteriori) significant difference of sound speeds for in-plane and bending modes. The explicit expression for the bending sound speed depending only on the graphene mass density, in-plane elastic constants and temperature is deduced as well as the characteristics of the microscopic corrugations of graphene. The obtained results are in good quantitative agreement with the data of real experiments and computer simulations.

BR2 reactor neutron beams

International Nuclear Information System (INIS)

Neve de Mevergnies, M.

1977-01-01

The use of reactor neutron beams is becoming increasingly more widespread for the study of some properties of condensed matter. It is mainly due to the unique properties of the ''thermal'' neutrons as regards wavelength, energy, magnetic moment and overall favorable ratio of scattering to absorption cross-sections. Besides these fundamental reasons, the impetus for using neutrons is also due to the existence of powerful research reactors (such as BR2) built mainly for nuclear engineering programs, but where a number of intense neutron beams are available at marginal cost. A brief introduction to the production of suitable neutron beams from a reactor is given. (author)

Resistive toroidal-field coils for tokamak reactors

International Nuclear Information System (INIS)

Kalnavarns, J.; Jassby, D.L.

1980-11-01

This paper analyzes the optimization of the geometry of resistive TF coils of rectangular bore for tokamak fusion test reactors and practical neutron generators. In examining the trade-offs between geometric parameters and magnetic field for reactors giving a specified neutron wall loading, either the resistive power loss or the lifetime coil cost can be minimized. Aspects of cooling, magnetic stress, and construction are addressed for several reference designs. Bending moment distributions in closed form have been derived for rectangular coils on the basis of the theory of rigid frames. Candidate methods of fabrication and of implementing demountable joints are summarized

AA, bending magnet, BLG

CERN Multimedia

CERN PhotoLab

1980-01-01

The very particular lattice of the AA required 2 types of dipole (bending magnets; BLG, long and narrow; BST, short and wide). The BLG had a steel length of 4.70 m, a good field width of 0.24 m, and a weight of about 70 t. Jean-Claude Brunet inspects the lower half of a BLG. For the BST magnets see 7811105 and 8006036.

Construction and operation of Clinch River Breeder Reactor Plant, docket no. 50-537, Oak Ridge, Roane County, Tennessee

International Nuclear Information System (INIS)

Anon.

1982-01-01

Construction and operation of the Clinch River Breeder Reactor Plant (CRBRP) in Oak Ridge, Tennessee are proposed. The CRBRP would use a liquid-sodium-cooled fast-breeder reactor to produce 975 megawatts of thermal energy (MWt) with the initial core loading of uranium- and plutonium-mixed oxide fuel. This heat would be transferred by heat exchangers to nonradioactive sodium in an intermediate loop and then to a steam cycle. A steam turbine generator would use the steam to produce 380 megawatts of electrical capacity (MWe). Future core design might result in gross power ratings of 1,121 MWt and 439 MWe. Exhaust steam from the turbine generator would be cooled in condensers using two mechanical draft cooling towers. The principal benefit would be the demonstration of the LMFBR concept for commercial use. Electricity generated would be a secondary benefit. Other impacts and effects are discussed

A dissipated energy comparison to evaluate fatigue resistance using 2-point bending

Directory of Open Access Journals (Sweden)

Cinzia Maggiore

2014-02-01

Full Text Available Fatigue is the main failure mode in pavement engineering. Typically, micro-cracks originate at the bottom of asphalt concrete layer due to horizontal tensile strains. Micro-cracks start to propagate towards the upper layers under repeated loading which can lead to pavement failure. Different methods are usually used to describe fatigue behavior in asphalt materials such as: phenomenological approach, fracture mechanics approach and dissipated energy approach. This paper presents a comparison of fatigue resistances calculated for different dissipated energy models using 2-point bending (2PB at IFSTTAR in Nantes. 2PB tests have been undertaken under different loading and environmental conditions in order to evaluate the properties of the mixtures (stiffness, dissipated energy, fatigue life and healing effect.

Formulation of Forming Load in V-Bending

Directory of Open Access Journals (Sweden)

Koumura Yuki

2016-01-01

Full Text Available A novel method is described to calculate the forming load in V-bending by a press brake. The data of forming load are collected by FEM analysis. With an increase of the punch stroke in V-bending, the forming load increases gradually after the elastic limit, and then decreases after showing the maximum value. The proposal formulation to trace the variations in the forming load curve includes the calculating method of the load of the elastic limit, the maximum load in air bending and the variations of the forming load before/after the bending stroke of the maximum load. The calculated precision is confirmed by comparing with the measured load-stroke curves in V-bending with a press brake.

Fission product release from SLOWPOKE-2 reactors

Energy Technology Data Exchange (ETDEWEB)

Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

1994-12-31

Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

Slice through an LHC bending magnet

CERN Multimedia

Slice through an LHC superconducting dipole (bending) magnet. The slice includes a cut through the magnet wiring (niobium titanium), the beampipe and the steel magnet yokes. Particle beams in the Large Hadron Collider (LHC) have the same energy as a high-speed train, squeezed ready for collision into a space narrower than a human hair. Huge forces are needed to control them. Dipole magnets (2 poles) are used to bend the paths of the protons around the 27 km ring. Quadrupole magnets (4 poles) focus the proton beams and squeeze them so that more particles collide when the beams’ paths cross. There are 1232 15m long dipole magnets in the LHC.

MANHATTAN PROJECT B REACTOR HANFORD WASHINGTON [HANFORD'S HISTORIC B REACTOR (12-PAGE BOOKLET)

Energy Technology Data Exchange (ETDEWEB)

GERBER MS

2009-04-28

The Hanford Site began as part of the United States Manhattan Project to research, test and build atomic weapons during World War II. The original 670-square mile Hanford Site, then known as the Hanford Engineer Works, was the last of three top-secret sites constructed in order to produce enriched uranium and plutonium for the world's first nuclear weapons. B Reactor, located about 45 miles northwest of Richland, Washington, is the world's first full-scale nuclear reactor. Not only was B Reactor a first-of-a-kind engineering structure, it was built and fully functional in just 11 months. Eventually, the shoreline of the Columbia River in southeastern Washington State held nine nuclear reactors at the height of Hanford's nuclear defense production during the Cold War era. The B Reactor was shut down in 1968. During the 1980's, the U.S. Department of Energy began removing B Reactor's support facilities. The reactor building, the river pumphouse and the reactor stack are the only facilities that remain. Today, the U.S. Department of Energy (DOE) Richland Operations Office offers escorted public access to B Reactor along a designated tour route. The National Park Service (NPS) is studying preservation and interpretation options for sites associated with the Manhattan Project. A draft is expected in summer 2009. A final report will recommend whether the B Reactor, along with other Manhattan Project facilities, should be preserved, and if so, what roles the DOE, the NPS and community partners will play in preservation and public education. In August 2008, the DOE announced plans to open B Reactor for additional public tours. Potential hazards still exist within the building. However, the approved tour route is safe for visitors and workers. DOE may open additional areas once it can assure public safety by mitigating hazards.

Effects of Bank Revetment on Sacramento River, California

Science.gov (United States)

Michael D. Harvey; Chester C. Watson

1989-01-01

Twelve low radius of curvature bends, half of which were rivetted, were studied in the Butte Basin reach of Sacramento River, California, to determine whether bank revetment deleteriously affected salmonid habitat. At low discharge (128.6 cubic meters/s) it was demonstrated that revetment does not cause channel narrowing or deepening, nor does it prevent re-entrainment...

Ketahanan Bending Komposit Hybrid Serat Batang Kelapa/Serat Gelas Dengan Matrik Urea Formaldehyde

Directory of Open Access Journals (Sweden)

Nasmi Herlina Sari

2012-11-01

Full Text Available The composite has its own advantages compared to other alternative techniques such material is strong, lightweight,corrosion-resistant, economical and so on. The purpose of this study was to investigate the characteristics of bending strengthfiber composite hybrid coconut trunk / fiber glass using urea formaldehyde resin.Hybrid palm trunk fiber /glass fiber composite have been made by hand lay up which volume fraction fiber hybridvariation namely 10:20, 15:15 and 20:10 (% with length fiber 2 cm. Every Tests conducted were bending testing with eachvariation performed three times repetition. Bending test specimens in accordance with standard ASTMD 790.The results of bending strength of palm trunk fiber hybrid composite / fiber-glass with random fiber direction that thehighest bending strength in the palm trunk fiber volume fraction 10% and 20% glass fiber that is 22.7 N/mm2.

Permanent bending and alignment of ZnO nanowires

Energy Technology Data Exchange (ETDEWEB)

Borschel, Christian; Spindler, Susann; Oertel, Michael; Ronning, Carsten [Institut fuer Festkoerperphysik, Friedrich-Schiller-Universitaet Jena, Max-Wien-Platz 1, 07743 Jena (Germany); Lerose, Damiana [MPI fuer Mikrostrukturphysik, Weinberg 2, 06120 Halle/Saale (Germany); Institut fuer Photonische Technologien, Albert-Einstein-Strasse 9, 07745 Jena (Germany); Bochmann, Arne [Institut fuer Photonische Technologien, Albert-Einstein-Strasse 9, 07745 Jena (Germany); Christiansen, Silke H. [Institut fuer Photonische Technologien, Albert-Einstein-Strasse 9, 07745 Jena (Germany); MPI fuer die Physik des Lichts, Guenther-Scharowsky-Str. 1, 91058 Erlangen (Germany); Nietzsche, Sandor [Zentrum fuer Elektronenmikroskopie, Friedrich-Schiller-Universitaet Jena, Ziegelmuehlenweg 1, 07743 Jena (Germany)

2011-07-01

Ion beams can be used to bend or re-align nanowires permanently, after they have been grown. We have irradiated ZnO nanowires with ions of different species and energy, achieving bending and alignment in various directions. We study the bending of single nanowires as well as the simultaneous alignment of large ensembles of ZnO nanowires in detail. Computer simulations show that the bending is initiated by ion beam induced damage. Dislocations are identified to relax stresses and make the bending and alignment permanent and resistant against annealing procedures.

The Maple reactor project

International Nuclear Information System (INIS)

Malkoske, G.R.; Labrie, J.-P.

2003-01-01

MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

78 FR 53482 - Entergy Operations, Inc., River Bend Station, Unit 1; Exemption

Science.gov (United States)

2013-08-29

... effect on the quality of the human environment (78 FR 50454; August 19, 2013). This exemption is... Regulatory Commission. Michele G. Evans, Director, Division of Operating Reactor Licensing, Office of Nuclear...

Bend-resistant large mode area fiber with novel segmented cladding

Science.gov (United States)

Ma, Shaoshuo; Ning, Tigang; Pei, Li; Li, Jing; Zheng, Jingjing

2018-01-01

A novel structure of segment cladding fiber (SCF) with characteristics of bend-resistance and large-mode-area (LMA) is proposed. In this new structure, the high refractive index (RI) core is periodically surrounded by high RI fan-segmented claddings. Numerical investigations show that effective single-mode operation of the proposed fiber with mode field area of 700 μm2 can be achieved when the bending radius is 15 cm. Besides, this fiber is insensitive to the bending orientation at the ranging of [-180°, 180°]. The proposed design shows great potential in high power fiber lasers and amplifiers with compact structure.

Examination of a failed reactor coolant pump rotating assembly from Crystal River Unit 3

International Nuclear Information System (INIS)

Hayner, G.O.; Lubnow, T.; Clary, M.

1990-01-01

On January 18, 1989, the A reactor coolant pump rotating assembly at the Crystal River Unit 3 Nuclear Power Plant failed during operation. A rotating assembly from this pump had previously failed in 1986. The reactor coolant pump was fabricated by Byron Jackson Pump Division of Borg-Warner Ind. Products, Inc. from UNS S66286 superalloy (Alloy A286). A root cause failure analysis examination was performed on the pump shaft and other components. The failure analysis included shaft vibrational mode and stress analyses, pump clearance and alignment analyses, and detailed destructive examination of the shaft and hydrostatic bearing assemblies. Based on the detailed physical examination of the shaft it was concluded that cracks initiated in the pump shaft at two sites approximately 180 0 apart in a band of shallow, thermally induced fatigue cracks. The cracks initiated at the bottom edge of the motor end shrink fit pad under the shrink fit sleeve supporting the hydrostatic bearing journal. The band of thermally induced fatigue cracks was apparently caused by mixing of cold seal injection water and hot reactor coolant in gaps between the pump shaft and sleeve. The motor end shrink fit was apparently not effective in preventing introduction of the seal injection water to this area. Initial crack propagation occurred by fatigue due to lateral vibration; however, the majority of crack propagation occurred by abnormal torsional fatigue loading induced by contact and sticking between the rotating and stationary portions of the hydrostatic bearing. Final fracture of the shaft occurred by torsional overload. Metallurgical characteristics and mechanical properties of the shaft were within design specification and probably did not significantly influence the cracking process

Modeling and Calculation of Dent Based on Pipeline Bending Strain

Directory of Open Access Journals (Sweden)

Qingshan Feng

2016-01-01

Full Text Available The bending strain of long-distance oil and gas pipelines can be calculated by the in-line inspection tool which used inertial measurement unit (IMU. The bending strain is used to evaluate the strain and displacement of the pipeline. During the bending strain inspection, the dent existing in the pipeline can affect the bending strain data as well. This paper presents a novel method to model and calculate the pipeline dent based on the bending strain. The technique takes inertial mapping data from in-line inspection and calculates depth of dent in the pipeline using Bayesian statistical theory and neural network. To verify accuracy of the proposed method, an in-line inspection tool is used to inspect pipeline to gather data. The calculation of dent shows the method is accurate for the dent, and the mean relative error is 2.44%. The new method provides not only strain of the pipeline dent but also the depth of dent. It is more benefit for integrity management of pipeline for the safety of the pipeline.

Hydrodynamic simulations of physical aquatic habitat availability for Pallid Sturgeon in the Lower Missouri River, at Yankton, South Dakota, Kenslers Bend, Nebraska, Little Sioux, Iowa, and Miami, Missouri, 2006-07

Science.gov (United States)

Jacobson, Robert B.; Johnson, Harold E.; Dietsch, Benjamin J.

2009-01-01

The objective of this study was to assess the sensitivity of habitat availability in the Lower Missouri River to discharge variation, with emphasis on habitats that might support spawning of the endangered pallid sturgeon. We constructed computational hydrodynamic models for four reaches that were selected because of evidence that sturgeon have spawned in them. The reaches are located at Miami, Missouri (river mile 259.6–263.5), Little Sioux, Iowa (river mile 669.6–673.5), Kenslers Bend, Nebraska (river mile 743.9–748.1), and Yankton, South Dakota reach (river mile 804.8–808.4). The models were calibrated for a range of measured flow conditions, and run for a range of discharges that might be affected by flow modifications from Gavins Point Dam. Model performance was assessed by comparing modeled and measured water velocities.A selection of derived habitat units was assessed for sensitivity to hydraulic input parameters (drag coefficient and lateral eddy viscosity). Overall, model results were minimally sensitive to varying eddy viscosity; varying lateral eddy viscosity by 20 percent resulted in maximum change in habitat units of 5.4 percent. Shallow-water habitat units were most sensitive to variation in drag coefficient with 42 percent change in unit area resulting from 20 percent change in the parameter value; however, no habitat unit value changed more than 10 percent for a 10 percent variation in drag coefficient. Sensitivity analysis provides guidance for selecting habitat metrics that maximize information content while minimizing model uncertainties.To assess model sensitivities arising from topographic variation from sediment transport on an annual time scale, we constructed separate models from two complete independent surveys in 2006 and 2007. The net topographic change was minimal at each site; the ratio of net topographic change to water volume in the reaches at 95 percent exceedance flow was less than 5 percent, indicating that on a reach

Strength tests of thin-walled elliptic duralumin cylinders in pure bending and in combined pure bending and torsion

Science.gov (United States)

Lundquist, Eugene E; Stowell, Elbridge Z

1942-01-01

An analysis is presented of the results of tests made by the Massachusetts Institute of Technology and by the National Advisory Committee for Aeronautics on an investigation of the strength of thin-walled circular and elliptic cylinders in pure bending and in combined torsion and bending. In each of the loading conditions, the bending moments were applied in the plane of the major axis of the ellipse.

Performance of a novel VUV bending magnet beamline

CERN Document Server

Song, Y F; Hsieh, T F; Huang, L R; Chung, S C; Cheng, N F; Hsiung, G Y; Wang, D J; Chen, C T; Tsang, K L

2001-01-01

A novel high resolution, high flux bending magnet beamline with an energy range from 5 to 40 eV has been constructed at SRRC. This Dragon-like beamline, which horizontally collects 50 mrad of synchrotron radiation from a bending magnet source, uses four cylindrical gratings with an included angle of 140 deg. and a movable curved exit slit. The average photon flux with an energy resolving power of 1000 is about 2x10 sup 1 sup 2 photons/s, which is among the highest of all existing VUV bending magnet beamlines. An energy resolving power of 24,000 at 6.8 eV has been obtained from the Schumann-Runge bands (B sup 3 limit construction operator in a limit construction/sum L: summation operator operator End lower limit of a limit construction u lower limit End limit End sup - /leftarrow/gets A: =leftward arrow X sup 3 limit construction operator in a limit construction/sum L: summation operator operator End lower limit of a limit construction g lower limit End limit End sup -) absorption spectra of O sub 2 gas. A pho...

Theory of bending waves with applications to disk galaxies

International Nuclear Information System (INIS)

Mark, J.W.K.

1982-01-01

A theory of bending waves is surveyed which provides an explanation for the required amplification of the warp in the Milky Way. It also provides for self-generated warps in isolated external galaxies. The shape of observed warps and partly their existence in isolated galaxies are indicative of substantial spheroidal components. The theory also provides a plausible explanation for the bending of the inner disk (<2 kpc) of the Milky Way

Pressurized water reactor simulator. Workshop material. 2. ed

International Nuclear Information System (INIS)

2005-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

Cernavoda NPP Unit 1: Ensuring heat sink at very low Danube river levels

International Nuclear Information System (INIS)

Urjan, D.

2005-01-01

Full text: On August 24, 2003 the summer heat wave has caused the Danube River to drop to its lowest level in more than a century, forcing a government commission of experts and a team of technical specialists from Cernavoda NPP to close Romania's unique nuclear reactor in operation at Cernavoda. The paper presents some of the required actions needed for plant shutdown and ensuring adequate fuel cooling at very low suction bay levels, due to the Danube River level drop (Danube waters cools the reactor). The water level in the Danube River at Cernavoda village, where the reactor is located, fell to a depth of less than three meters (10 feet) on Saturday, down from its usual level of almost seven meters (23 feet). Consequently, the Unit 1 nuclear power plant was shut down Sunday due to this record drought, which left insufficient water to cool down the reactor. Operating Instruction procedures were elaborated in order to provide a logical sequence of actions when the bay level decreases under 2.25 m, or the estimated level after 3 days will be lower than 1.8 m. When Raw Service Water (RSW) is lost, Recirculated Cooling Water (RCW) will remain in service only for Moderator, ESC, HT Pumps, HT Purification, D/C Cooler, LAC's, and D 2 O Feed Pump. Alternate water sources, like potable water and water from the fire protection system were taken in consideration in order to ensure heat sink to the RCW loads. At the same time, in case of total loss of Class III and Class IV Power, and Stand-By Diesel Generators unavailable because of the loss of heat sink provided by the RSW, Emergency Power System (EPS) was configured to supply directly the Class III Power 6 kV bus (BUG bus). Economical Impact According to a report, closing the nuclear plant costs Romania $500,000 a day. The total cost includes also losses due to a 40 percent reduction in hydroelectric power generation due to reduced river flow. The country had to cease power exports until the reactor comes back on line

Bending characteristics of resin concretes

Directory of Open Access Journals (Sweden)

Ribeiro Maria Cristina Santos

2003-01-01

Full Text Available In this research work the influence of composition and curing conditions in bending strength of polyester and epoxy concrete is analyzed. Various mixtures of resin and aggregates were considered in view of an optimal combination. The Taguchi methodology was applied in order to reduce the number of tests, and in order to evaluate the influence of various parameters in concrete properties. This methodology is very useful for the planning of experiments. Test results, analyzed by this methodology, shown that the most significant factors affecting bending strength properties of resin concretes are the type of resin, resin content and charge content. An optimal formulation leading to a maximum bending strength was achieved in terms of material parameters.

Neutron-physical characteristics of the TVRM-100 reactor with ten ring fuel channels

International Nuclear Information System (INIS)

Mikhajlov, V.M.; Myrtsymova, L.A.

1988-01-01

Three-dimensional heterogeneous calculations of TVRM-100 reactor which is a research reactor using enriched fuel with heavy-water moderator, coolant and reflector, are conducted. Achievable burnup depths depending on the number of removable FAs are presented. The maximum non-perturbed thermal neutron flux in the reflector is (2-1.8)x10 15 cm -2 c -1 ; mean flux on the fuel is 2.9x10 14 cm -2 c -1 . Energy release radial non-uniformity is 0.67, maximum bending by FA is ∼3.7. Reactivity temperature effect is negative and is equal to - 0.9x10 -4 grad -1 without accounting for experimental channels. Control rod efficiency in the radial reflector is high, but their location dose to experimental devices in the high neutron flux area is undesirable. 4 refs.; 5 figs

Limit load solutions for piping branch junctions under out-of-plane bending

International Nuclear Information System (INIS)

Xu, Ying Hu; Lee, Kuk Hee; Jeon, Jun Young; Kim, Yun Jae

2009-01-01

Approximate plastic limit load solutions for piping branch junctions under out-of plane bending are obtained from detailed three-dimensional (3-D) FE limit analyses based on elastic-perfectly plastic materials with the small geometry change option. Two types of bending are considered; out-of-plane bending to the branch pipe and out-of-plane bending to the run pipe. Accordingly closed-form approximations are proposed for piping branch junctions under out-of-plane bending based on the FE results. The proposed solutions are valid for the branch-to-run pipe radius and thickness from 0.0 to 1.0, and the mean radius-to-thickness ratio of the run pipe from 2.0 to 20.0. And, this study provides effects of reinforcement area on plastic limit loads.

Bending-Tolerant Anodes for Lithium-Metal Batteries.

Science.gov (United States)

Wang, Aoxuan; Tang, Shan; Kong, Debin; Liu, Shan; Chiou, Kevin; Zhi, Linjie; Huang, Jiaxing; Xia, Yong-Yao; Luo, Jiayan

2018-01-01

Bendable energy-storage systems with high energy density are demanded for conformal electronics. Lithium-metal batteries including lithium-sulfur and lithium-oxygen cells have much higher theoretical energy density than lithium-ion batteries. Reckoned as the ideal anode, however, Li has many challenges when directly used, especially its tendency to form dendrite. Under bending conditions, the Li-dendrite growth can be further aggravated due to bending-induced local plastic deformation and Li-filaments pulverization. Here, the Li-metal anodes are made bending tolerant by integrating Li into bendable scaffolds such as reduced graphene oxide (r-GO) films. In the composites, the bending stress is largely dissipated by the scaffolds. The scaffolds have increased available surface for homogeneous Li plating and minimize volume fluctuation of Li electrodes during cycling. Significantly improved cycling performance under bending conditions is achieved. With the bending-tolerant r-GO/Li-metal anode, bendable lithium-sulfur and lithium-oxygen batteries with long cycling stability are realized. A bendable integrated solar cell-battery system charged by light with stable output and a series connected bendable battery pack with higher voltage is also demonstrated. It is anticipated that this bending-tolerant anode can be combined with further electrolytes and cathodes to develop new bendable energy systems. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

SAVANNAH RIVER TECHNOLOGY CENTER MONTHLY REPORT AUGUST 1992

Energy Technology Data Exchange (ETDEWEB)

Ferrell, J.M.

1999-06-21

'This monthly report summarizes Programs and Accomplishments of the Savannah River Technology Center in support of activities at the Savannah River Site. The following categories are addressed: Reactor, Tritium, Separations, Environmental, Waste Management, General, and Items of Interest.'

Crack opening displacement of circumferential through-wall cracked cylinders subjected to tension and in-plane bending loads

International Nuclear Information System (INIS)

Yoo, Yeon-Sik

2003-01-01

This study is concerned with crack opening displacements (CODs) of cylinders with a circumferential through-crack which is subjected to tension and in-plane bending loads. Most studies about crack opening behavior have performed on membrane and global bending stresses. Moreover, they cannot be valid for large-scale structures. For simplicity on evaluation for structural integrity, crack opening displacement has been often calculated by plate or pipe model considering almost stresses as a membrane component. However, it is important to investigate ones close to real crack opening behaviors under stress states for reliability on evaluation. The results must be directly related to evaluate leakage detection in reactor vessel and the primary piping system of FBR structures. From that purpose, a series of FEM analyses were performed, and hence the characteristics of COD under an in-plane bending stress were compared with those under a membrane stress. In addition, the plate model was indicated to be unreasonable for application on large-scale pipes by comparing the plate model with the pipe model. The results of this study are expected to be valid for leakage evaluation of high temperature structures especially. (author)

Remote sensing analysis of thermal plumes at the Savannah River Plant

International Nuclear Information System (INIS)

Doak, E.L.

1985-01-01

The nuclear reactors of the Savannah River Plant (SRP) in Aiken, South Carolina, use cold water diverted from the Savannah River to dissipate unused thermal energy. This water is heated by heat exchangers of the reactors during the materials production process, and then returned to the natural drainage system. Thermal effluents were monitored by an airborne thermal infrared scanner during predawn overlights. Images were generated to show the surface temperature distribution of the thermal outfall plumes into the Savannah River. The thermal analysis provides information related to compliance with permit requirements of the regulatory agencies

The Spatial Structure of Planform Migration - Curvature Relation of Meandering Rivers

Science.gov (United States)

Guneralp, I.; Rhoads, B. L.

2005-12-01

Planform dynamics of meandering rivers have been of fundamental interest to fluvial geomorphologists and engineers because of the intriguing complexity of these dynamics, the role of planform change in floodplain development and landscape evolution, and the economic and social consequences of bank erosion and channel migration. Improved understanding of the complex spatial structure of planform change and capacity to predict these changes are important for effective stream management, engineering and restoration. The planform characteristics of a meandering river channel are integral to its planform dynamics. Active meandering rivers continually change their positions and shapes as a consequence of hydraulic forces exerted on the channel banks and bed, but as the banks and bed change through sediment transport, so do the hydraulic forces. Thus far, this complex feedback between form and process is incompletely understood, despite the fact that the characteristics and the dynamics of meandering rivers have been studied extensively. Current theoretical models aimed at predicting planform dynamics relate rates of meander migration to local and upstream planform curvature where weighting of the influence of curvature on migration rate decays exponentially over distance. This theoretical relation, however, has not been rigorously evaluated empirically. Furthermore, although models based on exponential-weighting of curvature effects yield fairly realistic predictions of meander migration, such models are incapable of reproducing complex forms of bend development, such as double heading or compound looping. This study presents the development of a new methodology based on parametric cubic spline interpolation for the characterization of channel planform and the planform curvature of meandering rivers. The use of continuous mathematical functions overcomes the reliance on bend-averaged values or piece-wise discrete approximations of planform curvature - a major limitation

Flow patterns and hydraulic losses in quasi-coil pipes : The effects of configuration of bend cross section, curvature ratio and bend angle

OpenAIRE

Shimizu, Yukimaru; Sugino, Koichi; Yasui, Masaji; Hayakawa, Yukitaka; Kuzuhara, Sadao

1985-01-01

Pipes with bend combinations are much used in the heat exchangers, since the curved path in the bends promotes the mixing in flow for active heat transfer. In the present paper, one of the pipes with bend combinations, namely, quasi-coiled pipes composed of many bend elements are investigated, and the relationships between the hydraulic loss and the secondary flow are studied experimentally. The configurations of the cross sections, the bent angles and the curvature ratios of the bend element...

Achieving the timely receipt of foreign research reactor spent nuclear fuel at the Savannah River site

International Nuclear Information System (INIS)

Brizes, C.M.; Clark, W.D; Thomas, J.; Andes, T.

1998-01-01

The May 1996 Record of Decision on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel states that the United States will accept spent nuclear fuel containing uranium of U.S.-origin from foreign research reactors through the year 2009. The best information available indicates that approximately 13,000 assemblies of Material Test Reactor (MTR) spent nuclear fuel from 29 countries are expected to be shipped to the Savannah River Site during the 13 years of the program. As of July 1998, 1,371 spent nuclear fuel assemblies from 12 foreign research reactors have been received at the SRS. That is, after more than two years of the FRR program (approximately 15 percent of the program time), 11 percent of the total assemblies have been received at SRS. Current projections show that most of the assemblies can be received by 2009, however if some of the eligible, non-participating countries decide to rejoin the program, a bottleneck would occur at the end of the program. Also adding to the potential for the bottleneck is a trend of shipments being moved out in the timeline. The Savannah River Site is working to be proactive in avoiding a bottleneck at the end of the program, but cooperation is required from all program participants to be successful. Activities currently in progress include inventory/information questionnaires, verifying fuel against cask(s) certificate of compliance (C. of C.), and collecting Appendix A information well in advance of shipping the SNF. The inventory/information sheets have been distributed to a select number of reactor facilities in the past, but work is in progress to refine the process. Information requested in the questionnaire includes inventory numbers, preferred shipping dates, and cask preferences. This information allows for improved shipment planning and helps to ensure that we are working to meet the needs of the reactor facilities. Current plans are to send the questionnaires to

Reliability evaluation of the Savannah River reactor leak detection system

International Nuclear Information System (INIS)

Daugherty, W.L.; Sindelar, R.L.; Wallace, I.T.

1991-01-01

The Savannah River Reactors have been in operation since the mid-1950's. The primary degradation mode for the primary coolant loop piping is intergranular stress corrosion cracking. The leak-before-break (LBB) capability of the primary system piping has been demonstrated as part of an overall structural integrity evaluation. One element of the LBB analyses is a reliability evaluation of the leak detection system. The most sensitive element of the leak detection system is the airborne tritium monitors. The presence of small amounts of tritium in the heavy water coolant provide the basis for a very sensitive system of leak detection. The reliability of the tritium monitors to properly identify a crack leaking at a rate of either 50 or 300 lb/day (0.004 or 0.023 gpm, respectively) has been characterized. These leak rates correspond to action points for which specific operator actions are required. High reliability has been demonstrated using standard fault tree techniques. The probability of not detecting a leak within an assumed mission time of 24 hours is estimated to be approximately 5 x 10 -5 per demand. This result is obtained for both leak rates considered. The methodology and assumptions used to obtain this result are described in this paper. 3 refs., 1 fig., 1 tab

Reactor operation environmental information document

Energy Technology Data Exchange (ETDEWEB)

Bauer, L.R.; Hayes, D.W.; Hunter, C.H.; Marter, W.L.; Moyer, R.A.

1989-12-01

This volume is a reactor operation environmental information document for the Savannah River Plant. Topics include meteorology, surface hydrology, transport, environmental impacts, and radiation effects. 48 figs., 56 tabs. (KD)

Symmetric bends how to join two lengths of cord

CERN Document Server

Miles, Roger E

1995-01-01

A bend is a knot securely joining together two lengths of cord (or string or rope), thereby yielding a single longer length. There are many possible different bends, and a natural question that has probably occurred to many is: "Is there a 'best' bend and, if so, what is it?"Most of the well-known bends happen to be symmetric - that is, the two constituent cords within the bend have the same geometric shape and size, and interrelationship with the other. Such 'symmetric bends' have great beauty, especially when the two cords bear different colours. Moreover, they have the practical advantage o

Tunable waveguide bends with graphene-based anisotropic metamaterials

KAUST Repository

Chen, Zhao-xian; Chen, Ze-guo; Ming, Yang; Wu, Ying; Lu, Yan-qing

2016-01-01

We design tunable waveguide bends filled with graphene-based anisotropic metamaterials to achieve a nearly perfect bending effect. The anisotropic properties of the metamaterials can be described by the effective medium theory. The nearly perfect bending effect is demonstrated by finite element simulations of various structures with different bending curvatures and shapes. This effect is attributed to zero effective permittivity along the direction of propagation and matched effective impedance at the interfaces between the bending part and the dielectric waveguides. We envisage that the design will be applicable in the far-infrared and terahertz frequency ranges owing to the tunable dielectric responses of graphene.

Tunable waveguide bends with graphene-based anisotropic metamaterials

KAUST Repository

Chen, Zhao-xian

2016-01-15

We design tunable waveguide bends filled with graphene-based anisotropic metamaterials to achieve a nearly perfect bending effect. The anisotropic properties of the metamaterials can be described by the effective medium theory. The nearly perfect bending effect is demonstrated by finite element simulations of various structures with different bending curvatures and shapes. This effect is attributed to zero effective permittivity along the direction of propagation and matched effective impedance at the interfaces between the bending part and the dielectric waveguides. We envisage that the design will be applicable in the far-infrared and terahertz frequency ranges owing to the tunable dielectric responses of graphene.

Bending stresses in Facetted Glass Shells

DEFF Research Database (Denmark)

Bagger, Anne; Jönsson, Jeppe; Almegaard, Henrik

2008-01-01

A shell structure of glass combines a highly effective structural principle with a material of optimal permeability to light. A facetted shell structure has a piecewise plane geometry, and together the facets form an approximation to a curved surface. A distributed load on a plane-based facetted...... structure will locally cause bending moments in the loaded facets. The bending stresses are dependent on the stiffness of the joints. Approximate solutions are developed to estimate the magnitude of the bending stresses. A FE-model of a facetted glass shell structure is used to validate the expressions...

Processing test of an upgraded mechanical design for PERMCAT reactor

International Nuclear Information System (INIS)

Borgognoni, Fabio; Demange, David; Doerr, Lothar; Tosti, Silvano; Welte, Stefan

2010-01-01

The PERMCAT membrane reactor is a coaxial combination of a Pd/Ag permeator membrane and a catalyst bed. This device has been proposed for processing fusion reactor plasma exhaust gas. A stream containing tritium (up to 1% of tritium in different chemical forms such as water, methane or molecular hydrogen) is decontaminated in the PERMCAT by counter-current isotopic swamping with protium. Different mechanical designs of the membrane reactor have been proposed to improve robustness and lifetime. The ENEA membrane reactor uses a permeator tube with a length of about 500 mm produced via cold-rolling and diffusion welding of Pd/Ag thin foils: two stainless steel pre-tensioned bellows have been applied to the Pd/Ag tube in order to avoid any significant compressive and bending stresses due to the permeator tube elongation consequent to the hydrogen uptake. An experimental test campaign has been performed using this reactor in order to assess the influence of different operating parameters and to evaluate the overall performance (decontamination factor). Tests have been carried out on two reactor prototypes: a defect-free membrane with complete (infinite) hydrogen selectivity and not perm-selective membrane. In this last case, the study has been aimed at verifying the behaviour of the PERMCAT devices under non-normal (accidental) conditions in the view of providing information for future safety analysis. The paper will present the specific mechanical design and the experimental results of tests based on isotopic exchange between H 2 O and D 2 .

Effects of laser bending on the microstructural constituents

CSIR Research Space (South Africa)

Tshabalala, L

2012-01-01

Full Text Available This article will illustrate the correlation between microstructural and microhardness changes in high-strength-low-alloy steel that occur as a result of laser-bending. Laser bending is a process of bending metal shapes using the laser beam...

Effects of large bending deflections on blade flutter limits

Energy Technology Data Exchange (ETDEWEB)

Kallesoee, Bjarne Skovmose; Hartvig Hansen, Morten

2008-04-15

The coupling of bending and torsion due to large blade bending are assumed to have some effects of the flutter limits of wind turbines. In the present report, the aeroelastic blade model suggested by Kallesoee, which is similar to a second order model, is used to investigate the aeroelastic stability limits of the RWT blade with and without the effects of the large blade deflection. The investigation shows no significant change of the flutter limit on the rotor speed due to the blade deflection,whereas the first edgewise bending mode becomes negatively damped due to the coupling with blade torsion which causes a change of the effective direction of blade vibration. These observations are confirmed by nonlinear aeroelastic simulations using HAWC2. This work is part of the UpWind project funded by the European Commission under the contract number SES6-CT-2005-019945 which is gratefully acknowledged. This report is the deliverable D2.3 of the UpWind project. (au)

Infrared spectra and tunneling dynamics of the N2-D2O and OC-D2O complexes in the v2 bend region of D2O.

Science.gov (United States)

Zhu, Yu; Zheng, Rui; Li, Song; Yang, Yu; Duan, Chuanxi

2013-12-07

The rovibrational spectra of the N2-D2O and OC-D2O complexes in the v2 bend region of D2O have been measured in a supersonic slit jet expansion using a rapid-scan tunable diode laser spectrometer. Both a-type and b-type transitions were observed for these two complexes. All transitions are doubled, due to the heavy water tunneling within the complexes. Assuming the tunneling splittings are the same in K(a) = 0 and K(a) = 1, the band origins, all three rotational and several distortion constants of each tunneling state were determined for N2-D2O in the ground and excited vibrational states, and for OC-D2O in the excited vibrational state, respectively. The averaged band origin of OC-D2O is blueshifted by 2.241 cm(-1) from that of the v2 band of the D2O monomer, compared with 1.247 cm(-1) for N2-D2O. The tunneling splitting of N2-D2O in the ground state is 0.16359(28) cm(-1), which is about five times that of OC-D2O. The tunneling splittings decrease by about 26% for N2-D2O and 23% for OC-D2O, respectively, upon excitation of the D2O bending vibration, indicating an increase of the tunneling barrier in the excited vibrational state. The tunneling splittings are found to have a strong dependence on intramolecular vibrational excitation as well as a weak dependence on quantum number K(a).

Advances in Reactor Physics, Mathematics and Computation. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.

Savannah River Site peer evaluator standards: Operator assessment for restart

International Nuclear Information System (INIS)

1990-01-01

Savannah River Site has implemented a Peer Evaluator program for the assessment of certified Central Control Room Operators, Central Control Room Supervisors and Shift Technical Engineers prior to restart. This program is modeled after the nuclear Regulatory Commission's (NRC's) Examiner Standard, ES-601, for the requalification of licensed operators in the commercial utility industry. It has been tailored to reflect the unique differences between Savannah River production reactors and commercial power reactors

Electrostatic bending response of a charged helix

Science.gov (United States)

Zampetaki, A. V.; Stockhofe, J.; Schmelcher, P.

2018-04-01

We explore the electrostatic bending response of a chain of charged particles confined on a finite helical filament. We analyze how the energy difference Δ E between the bent and the unbent helical chain scales with the length of the helical segment and the radius of curvature and identify features that are not captured by the standard notion of the bending rigidity, normally used as a measure of bending tendency in the linear response regime. Using Δ E to characterize the bending response of the helical chain we identify two regimes with qualitatively different bending behaviors for the ground state configuration: the regime of small and the regime of large radius-to-pitch ratio, respectively. Within the former regime, Δ E changes smoothly with the variation of the system parameters. Of particular interest are its oscillations with the number of charged particles encountered for commensurate fillings which yield length-dependent oscillations in the preferred bending direction of the helical chain. We show that the origin of these oscillations is the nonuniformity of the charge distribution caused by the long-range character of the Coulomb interactions and the finite length of the helix. In the second regime of large values of the radius-to-pitch ratio, sudden changes in the ground state structure of the charges occur as the system parameters vary, leading to complex and discontinuous variations in the ground state bending response Δ E .

Effects of tanalith-e impregnation substance on bending strengths and modulus of elasticity in bending of some wood types

Directory of Open Access Journals (Sweden)

Hakan Keskin

2016-04-01

Full Text Available The aim of this study was to investigate the effects of impregnation with Tanalith-E on the bending strengths and modulus of elasticity in bending of some wood types. The test samples prepared from beech, oak, walnut, poplar, ash and pine wood materials - that are of common use in the forest products industry of TURKEY - according to TS 345, were treated with according to ASTM D 1413-76 substantially. Un-impregnated samples according to impregnated wood materials, the bending strengths in beech to 6.83%, 5.12% in ash, 5.93% in pine, the elasticity module values to 7.15% in oak and ash, at a rate of 6.58% in the higher were found. The highest values of bending strengths and modulus of elasticity in bending were obtained in beech and ash woods impregnated with Tanalith-E, whereas the lowest values were obtained in the poplar wood.

Compliance of the Savannah River Plant P-Reactor cooling system with environmental regulations. Demonstrations in accordance with Sections 316(a) and (b) of the Federal Water Pollution Control Act of 1972

International Nuclear Information System (INIS)

Wilde, E.W.

1985-12-01

This document presents demonstrations under Sections 316(a) and (b) of the Federal Water Pollution Control Act of 1972 for the P-Reactor cooling system at the Savannah River Plant (SRP). The demonstrations were mandated when the National Pollution Discharge Elimination System (NPDES) permit for the SRP was renewed and the compliance point for meeting South Carolina Class B water quality criteria in the P-Reactor cooling system was moved from below Par Pond to the reactor cooling water outfall, No. P-109. Extensive operating, environmental, and biological data, covering most of the current P-Reactor cooling system history from 1958 to the present are discussed. No significant adverse effects were attributed to the thermal effluent discharged to Par Pond or the pumping of cooling water from Par Pond to P Reactor. It was conluded that Par Pond, the principal reservoir in the cooling system for P Reactor, contains balanced indigenous biological communities that meet all criteria commonly used in defining such communities. Par Pond compares favorably with all types of reservoirs in South Carolina and with cooling lakes and reservoirs throughout the southeast in terms of balanced communities of phytoplankton, macrophytes, zooplankton, macroinvertebrates, fish, and other vertebrate wildlife. The report provides the basis for negotiations between the South Carolina Department of Health and Environmental Control (SCDHEC) and the Department of Energy - Savannah River (DOE-SR) to identify a mixing zone which would relocate the present compliance point for Class B water quality criteria for the P-Reactor cooling system

Draft environmental impact statement for the siting, construction, and operation of New Production Reactor capacity. Volume 2, Sections 1-6

Energy Technology Data Exchange (ETDEWEB)

1991-04-01

This (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site. The EIS programmatic alternatives address Departmental decisions to be made on whether to build new production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains the analysis of programmatic alternatives, project alternatives, affected environment of alternative sites, environmental consequences, and environmental regulations and permit requirements.

Microscopic fracture of filaments and its relation to the critical current under bending deformation in (Bi,Pb)2Sr2Ca2Cu3O10 composite superconducting tapes

International Nuclear Information System (INIS)

Hojo, Masaki; Nakamura, Mitsuhiro; Matsuoka, Tomoe; Tanaka, Mototsugu; Ochiai, Shojiro; Sugano, Michinaka; Osamura, Kozo

2003-01-01

The strain dependence of the critical current, I c , of (Bi,Pb) 2 Sr 2 Ca 2 Cu 3 O 10 (Bi2223)/Ag/Ag-Mg composite superconducting tapes has been studied both experimentally and analytically under bending deformation. Tests have been carried out for one type of tape used in the VAMAS bending round-robin programme. The complex stress-strain behaviour of each component was first analysed in tension. This was done by comparing the stress-strain curves of composite tapes with those of Ag and Ag-Mg alloy tapes. Here, the plastic deformation (work hardening) of Ag and Ag-Mg alloy, and the thermal residual strain due to the manufacturing process were taken into account. The fracture strain of Bi2223 filaments was inversely determined as 0.08% to meet the global tensile stress-strain curve of the composite tape. The calculated stress-strain curves finally agreed well with the experimental results when the as-supplied bending strain was taken into account. Then, the analysis was modified to fit the bending deformation. Here, the movement of the neutral axis due to the non-symmetric and elastic-plastic stress-strain curves of the components and their Bauschinger effect were taken into account. The relative decrease of I c with the increase in the Bi2223 tape curvature was calculated from the volume fraction of the broken filaments. The calculated I c agreed well with the experimental results when the movement of the neutral axis and the Bauschinger effect were taken into account. Microscopic observation of the spatial distribution of the filament fracture indicated that the damage occurred at the outermost layer on the tensile side when the curvature was small, and then the damage front shifted to the inside layers. The observed fracture behaviour of the Bi2223 filament agreed well with the estimated location based on the above analysis

Siting of research reactors

International Nuclear Information System (INIS)

1987-01-01

The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

Metal-bending brake facilitates lightweight, close-tolerance fabrication

Science.gov (United States)

Ercoline, A. L.; Wilton, K. B.

1964-01-01

A lightweight, metal bending brake ensures very accurate bends. Features of the brake that adapt it for making complex reverse bends to close tolerances are a pronounced relief or cutaway of the underside of the bodyplate combined with modification in the leaf design and its suspension.

Bending strength of glass-ceramics based on 3CaO.P2O5-SiO2-MgO glass system

International Nuclear Information System (INIS)

Daguano, J.K.M.F.; Suzuki, P.A.; Santos, C.; Fernandes, M.H.V.; Elias, C.N.

2009-01-01

In this work, the Modulus of Rupture of bioactive glass-ceramic based on 3CaO.P 2 O 5 -SiO 2 -MgO system was investigated, aiming its use in bone-restorations. The mechanical property was correlated with microstructural and crystallographic features of this material. High-purity starting-powders, CaCO 3 , SiO 2 , MgO, Ca (H 2 PO 4 ).H 2 O, were used in this study. The powders were mixed in a stoichiometric ratio, using planetary ball-mill. The suspensions were dried, sieved and melted at 1600 deg C, for 4h. The casting ones were cooled quickly until annealing temperature 700 deg C, in which remained for 2h, with controlled cooling-rate until ambient temperature. Bulks of glass were heat-treated with temperatures varying between 700 deg C and 1100 deg C, for 4h, being after that, cooled at 3 deg C/min. Bioactive glass and glass-ceramic were characterized by HRXRD (high resolution X-ray diffraction), where whitlockite was main phase. The microstructure was analyzed by scanning electronic microscopy. Modulus of Rupture was determined by four-point bending testing using specimens of 1.5 x 2 x 25 mm and glasses presented strength near to 70MPa, while glass ceramics treated at 975 deg C-4h, presented bending strength of 120MPa. (author)

Load tests with a pipe bend DN 425, applying slowly changing bending loads up to occurrence of leak

International Nuclear Information System (INIS)

Uhlmann, D.; Hunger, H.

1990-01-01

The experimental program deals with the formation of incipient cracks and subsequent crack growth of axially oriented cracks at a pipe bend with a nominal width of DN 425. The pipe bend consists of the ferritic material 20MnMoNi55. The numerical experiments by means of 3 D-FE analyses concentrate on determining the influence of the asymmetric crack depths at the two bend halves, and of the multiple crack fields, on the effective crack strain. (DG) [de

Working Group 2 summary: Space charge effects in bending systems

International Nuclear Information System (INIS)

Bohn, C.L.; Emma, P.J.

2000-01-01

At the start of the Workshop, the authors asked the Working Group 2 participants to concentrate on three basic goals: (1) survey the status of how comprehensively the physics concerning space-charge effects in bends is understood and how complete is the available ensemble of analytic and computational tools; (2) guided by data from experiments and operational experience, identify sources of, and cures for, beam degradation; and (3) review space-charge physics in rings and the limitations it introduces. As the Workshop unfolded, the third goal naturally folded into the other two goals, and these goals, they believe, were fulfilled in that the Working Group was able to compile an end product consisting of a set of recommendations for potentially fruitful future work. This summary constitutes an overview of the deliberations of the Working Group, and it is their hope that the summary clarifies the motivation for the recommended work listed at the end. The summary is organized according to the two aforementioned goals, and the prime topics of discussion appear as subsections under these goals

A Novel Rotary Piezoelectric Motor Using First Bending Hybrid Transducers

Directory of Open Access Journals (Sweden)

Yingxiang Liu

2015-08-01

Full Text Available We report a novel rotary piezoelectric motor using bending transducers in this work. Three transducers are used to drive a disk-shaped rotor together by the elliptical movements of their driving tips; these motions are produced by the hybrid of two first bending vibration modes. The proposed piezoelectric transducer has a simple structure as it only contains an aluminum alloy beam and four pieces of PZT plates. Symmetrical structure is the only necessary condition in the design process as it will ensure the resonance frequencies of the two orthogonal first bending modes are equal. Transducers with first bending resonance frequency of about 53 kHz were fabricated and assembled into a rotary motor. The proposed motor exhibits good performance on speed and torque control. Under a working frequency of 53.2 kHz, the maximum no-load speed and the maximum torque of the prototype are tested to be 53.3 rpm and of 27 mN·m.

Reactor operation environmental information document

Energy Technology Data Exchange (ETDEWEB)

Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

1989-12-01

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

Contribution to descriptions of pressurized water reactors close to Danish territory

International Nuclear Information System (INIS)

Johansson, T.

1993-04-01

This paper is part of a report describing Pressurrized Water Reactors (PWR's) close to Danish territory. The full report is the outcome of a working group formed as part of a continued collaboration between the Department of Electrophysics, DTU, and the Risoe National Laboratory, a collaboration with the purpose of maintaining Danish knowledge on commercial nuclear power plants. The reactor dealt with in this report are the Ringhals, 2, 3 and 4 reactors, situated 50 km south of Goeteborg on the western coast of Sweden, and the reactors in Stade and Brokdorf in the northern part of Germany downstream from Hamburg at the river Elben. This paper deals with the following subjects for all the above mentioned reactors: Reactor core and other vessel components. Reactivity control systems and Fuel and component handling and storage systems. (EG)

Internal structure of reactor building for Madras Atomic Power Project

International Nuclear Information System (INIS)

Pandit, D.P.

1975-01-01

The structural configuration and analysis of structural elements of the internal structure of reactor building for the Madras Atomic Power Project has been presented. Two methods of analysis of the internal structure, viz. Equivalent Plane Frame and Finite Element Method, are explained and compared with the use of bending moments obtained. (author)

Mechanical behaviour of bending bucky-gel actuators and its representation

International Nuclear Information System (INIS)

Kruusamäe, Karl; Mukai, Ken; Sugino, Takushi; Asaka, Kinji

2014-01-01

Bucky-gel actuators are ionic electromechanically active materials that bend in response to a low-voltage excitation. While bending actuators may offer new approaches in engineering solutions, the characterization of bending poses many challenges in comparison to conventional rotary motion. It is often desired to reduce the bending behaviour to a single parameter, which may lead to the loss of accuracy in modelling. A high-speed laser profilometer is utilized to characterize the bending response of different bucky-gel actuators at their full length and to critically compare the applicability of existing representation tools for bending. The best analytical representation of the bending of a bucky-gel actuator is found to be in the form of a power function. It is also observed that, along the length of the actuator, sections closer to the electrical input clamp exhibit back-relaxation (a common drawback for bending ionic actuators) already when the far end of the bending strip is still in forward motion. (paper)

Bend-imitating theory and electron scattering in sharply-bent quantum nanowires

International Nuclear Information System (INIS)

Vakhnenko, O.O.

2011-01-01

The concept of bend-imitating description as applied to the one-electron quantum mechanics in sharply-bent ideal electron waveguides and its development into a self consistent theory are presented. In the framework of bend-imitating approach, the investigation of the electron scattering in a doubly-bent 2D quantum wire with S-like bend has been made, and the explicit dependences of the transmission and reflection coefficients on geometrical parameters of a structure, as well as on the electron energy, have been obtained. The total elimination of the mixing between the scattering channels of a S-like bent quantum wire is predicted.

Channel morphodynamics in four reaches of the Lower Missouri River, 2006-07

Science.gov (United States)

Elliott, Caroline M.; Reuter, Joanna M.; Jacobson, Robert B.

2009-01-01

Channel morphodynamics in response to flow modifications from Gavins Point Dam are examined in four reaches of the Lower Missouri River. Measures include changes in channel morphology and indicators of sediment transport in four 6 kilometer long reaches located downstream from Gavins Point Dam, near Yankton, South Dakota, Kenslers Bend, Nebraska, Little Sioux, Iowa, and Miami, Missouri. Each of the four reaches was divided into 300 transects with a 20-meter spacing and surveyed during the summer in 2006 and 2007. A subset of 30 transects was randomly selected and surveyed 7-10 times in 2006-07 over a wide range of discharges including managed and natural flow events. Hydroacoustic mapping used a survey-grade echosounder and a Real Time Kinematic Global Positioning System to evaluate channel change. Acoustic Doppler current profiler measurements were used to evaluate bed-sediment velocity. Results indicate varying amounts of deposition, erosion, net change, and sediment transport in the four Lower Missouri River reaches. The Yankton reach was the most stable over monthly and annual time-frames. The Kenslers Bend and Little Sioux reaches exhibited substantial amounts of deposition and erosion, although net change was generally low in both reaches. Total, or gross geomorphic change was greatest in the Kenslers Bend reach. The Miami reach exhibited varying rates of deposition and erosion, and low net change. The Yankton, Kenslers Bend, and Miami reaches experienced net erosion during the time period that bracketed the managed May 2006 spring rise event from Gavins Point Dam.

Safety re-assessment of AECL test and research reactors

International Nuclear Information System (INIS)

Winfield, D.J.

1990-01-01

Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

Study to compare the performance of two designs to prevent river bend erosion in Arctic environments.

Science.gov (United States)

2010-09-01

Messing with Mother Nature takes knowledge and work, and she is hard to outfox, especially when it comes to redirecting rivers. To : protect infrastructure, however, sometimes river flow must be altered. This study focuses on two erosion-control proj...

Symmetric tape round REBCO wire with J e (4.2 K, 15 T) beyond 450 A mm-2 at 15 mm bend radius: a viable candidate for future compact accelerator magnet applications

Science.gov (United States)

Kar, Soumen; Luo, Wenbo; Ben Yahia, Anis; Li, Xiaofen; Majkic, Goran; Selvamanickam, Venkat

2018-04-01

Round REBCO (RE = rare earth) wires of 1.6-1.85 mm diameter have been fabricated using ultrathin REBCO tapes where the superconductor film is positioned near the geometric center. Such symmetric tape round (STAR) wires exhibit excellent tolerance to bend strain with a critical current retention of more than 97% when bent to a radius of 15 mm. A 1.6 mm diameter REBCO STAR wire made with six 2.5 mm wide symmetric tapes reached an engineering current density (J e) of 454 A mm-2 at 4.2 K in a background field of 15 T at a bend radius of 15 mm. Such superior performance at a small bend radius can enable fabrication of future accelerator magnets, operating at magnetic fields above 20 T.

Calculation of Savannah River K Reactor Mark-22 assembly LOCA/ECS power limits

International Nuclear Information System (INIS)

Fischer, S.R.; Farman, R.F.; Birdsell, S.A.

1992-01-01

This paper summarizes the results of TRAC-PF1/MOD3 calculations of Mark-22 fuel assembly of loss-of-coolant accident/emergency cooling system (LOCA/ECS) power limits for the Savannah River Site (SRS) K Reactor. This effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to perform confirmatory power limits calculations for the SRS K Reactor. A method using a detailed three-dimensional (3D) TRAC model of the Mark-22 fuel assembly was developed to compute LOCA/ECS power limits. Assembly power was limited to ensure that no point on the fuel assembly walls would exceed the local saturation temperature. The detailed TRAC model for the Mark-22 assembly consisted of three concentric 3D vessel components which simulated the two targets, two fuel tubes, and three main flow channels of the fuel assembly. The model included 100% eccentricity between the assembly annuli and a 20% power tilt. Eccentricity in the radial alignment of the assembly annuli arises because axial spacer ribs that run the length of the fuel and targets are used. Wall-shear, interfacial-shear, and wall heat-transfer correlations were developed and implemented in TRAC-PF1/MOD3 specifically for modeling flow and heat transfer in the narrow ribbed annuli encountered in the Mark-22 fuel assembly design. We established the validity of these new constitutive models using separate-effects benchmarks. TRAC system calculations of K Reactor indicated that the limiting ECS-phase accident is a double-ended guillonite break in a process water line at the pump discharge (i.e., a PDLOCA). The fuel assembly with the minimum cooling potential is identified from this system calculation. Detailed assembly calculations then were performed using appropriate boundary conditions obtained from this limiting system LOCA. Coolant flow rates and pressure boundary conditions were obtained from this system calculation and applied to the detailed assembly model

Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

International Nuclear Information System (INIS)

Bussac, J.; Reuss, P.

1985-01-01

This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

Radioiodine in the Savannah River Site environment

Energy Technology Data Exchange (ETDEWEB)

Kantelo, M.V.; Bauer, L.R.; Marter, W.L.; Murphy, C.E. Jr.; Zeigler, C.C.

1993-01-15

Radioiodine, which is the collective term for all radioactive isotopes of the element iodine, is formed at the Savannah River Site (SRS) principally as a by-product of nuclear reactor operations. Part of the radioiodine is released to the environment during reactor and reprocessing operations at the site. The purpose of this report is to provide an introduction to radioiodine production and disposition, its status in the environment, and the radiation dose and health risks as a consequence of its release to the environment around the Savannah River Plant. A rigorous dose reconstruction study is to be completed by thee Center for Disease Control during the 1990s.

Radioiodine in the Savannah River Site environment

International Nuclear Information System (INIS)

Kantelo, M.V.; Bauer, L.R.; Marter, W.L.; Murphy, C.E. Jr.; Zeigler, C.C.

1993-01-01

Radioiodine, which is the collective term for all radioactive isotopes of the element iodine, is formed at the Savannah River Site (SRS) principally as a by-product of nuclear reactor operations. Part of the radioiodine is released to the environment during reactor and reprocessing operations at the site. The purpose of this report is to provide an introduction to radioiodine production and disposition, its status in the environment, and the radiation dose and health risks as a consequence of its release to the environment around the Savannah River Plant. A rigorous dose reconstruction study is to be completed by thee Center for Disease Control during the 1990s

High-sensitivity bend angle measurements using optical fiber gratings.

Science.gov (United States)

Rauf, Abdul; Zhao, Jianlin; Jiang, Biqiang

2013-07-20

We present a high-sensitivity and more flexible bend measurement method, which is based on the coupling of core mode to the cladding modes at the bending region in concatenation with optical fiber grating serving as band reflector. The characteristics of a bend sensing arm composed of bending region and optical fiber grating is examined for different configurations including single fiber Bragg grating (FBG), chirped FBG (CFBG), and double FBGs. The bend loss curves for coated, stripped, and etched sections of fiber in the bending region with FBG, CFBG, and double FBG are obtained experimentally. The effect of separation between bending region and optical fiber grating on loss is measured. The loss responses for single FBG and CFBG configurations are compared to discover the effectiveness for practical applications. It is demonstrated that the sensitivity of the double FBG scheme is twice that of the single FBG and CFBG configurations, and hence acts as sensitivity multiplier. The bend loss response for different fiber diameters obtained through etching in 40% hydrofluoric acid, is measured in double FBG scheme that resulted in a significant increase in the sensitivity, and reduction of dead-zone.

Responses of platinum, vanadium and cobalt self-powered flux detectors near simulated booster rods in a ZED-2 mockup of a Bruce reactor core

International Nuclear Information System (INIS)

French, P.M.; Shields, R.B.; Kroon, J.C.

1978-02-01

The static responses of Pt, V and Co self-powered detectors have been compared with copper-foil neutron activation profiles in reference and perturbed Bruce reactor core mockups assembled in the ZED-2 test reactor at Chalk River Nuclear Laboratories. The results indicate that the normalized response of each self-powered detector is an accurate measure of the thermal-neutron flux at locations greater than one lattice pitch from either a booster rod or the core boundary. They indicate that, in the Bruce booster/detector configuration, the normalized static Pt response overestimates the neutron flux by less than 3.5% upon full booster-rod insertion. (author)

Bending strain study of Bi-2223/Ag tapes using Hall sensor magnetometry

International Nuclear Information System (INIS)

Lahtinen, M.; Paasi, J.; Sarkaniemi, J.; Han, Z.; Freltoft, T.

1996-01-01

The influence of room temperature bending on critical current (I c ) of Bi-2223/Ag tapes is studied by Hall sensor magnetometry, four-point method and scanning electron microscopy. Hall sensor magnetometry allows one to assess tape homogeneity and the amount of mechanical damage caused by bending. The microstructure of the Bi-2223 ceramic is found to strongly affect the tape behavior under bending strain. In a tape with moderate I c = 6.1 A at 77 K and a porous ceramic core, crack propagation took place normal to the Ag-ceramic interface, whereas in tapes with dense core, I c above 10 A at 77 K, cracks propagated in the tape plane. In monofilamentary tapes core homogeneity correlated with good bending strain performance. In multifilamentary tapes crack propagation between filaments was prohibited by the Ag matrix, thus leading to enhanced strain tolerance. In the high I c tapes studied, bending to 25 mm radius resulted in 1%--2% I c degradation

Pilot study risk assessment for selected problems at the Savannah River Site (SRS)

International Nuclear Information System (INIS)

Hamilton, L.D.; Holtzman, S.; Meinhold, A.; Morris, S.C.; Pardi, R.; Sun, C.; Daniels, J.I.; Layton, D.; McKone, T.E.; Straume, T.; Anspaugh, L.

1993-03-01

An assessment of the health risks was made for releases of tritium and 137 Cs from the Savannah River Site (SRS) at water-receptor locations downriver. Although reactor operations were shut down at the SRS in 1989, liquid wastes continue to be released to the Savannah River either by direct discharges into onsite surface waters or by groundwater transport into surface waters from waste facilities. Existing state mandates will cause the liquid waste streams from future operations to go directly into surface waters. Two drinking water processing plants take water from the river approximately 129 km downriver from the SRS. Potential incremental risks of cancer fatality to individuals and each population were analyzed for either no further reactor operations or resumption of operation of one specific reactor

Alternatives to L startup: new production reactor

International Nuclear Information System (INIS)

Hostetler, D.E.

1983-01-01

An alternative to renewed operation of L Reactor for increased production of nuclear materials would be the construction and operation of a New Production Reactor (NPR). This report describes a conceptual design for a low temperature heavy water reactor with no electricity generation (LTHWR-NE) to be built as a new production reactor at the Savannah River Plant (SRP). The reactor design is based on the proven SRP reactor design with enhancements and state-of-the-art equipment. Aluminum cladding temperatures would be the same as with current operations. The power and productivity of the new reactor would be greater than L Reactor by about 30%. However, the estimated time from authorization to startup is 10 years. Thus an NPR could not contribute to material production until late 1993 at the earliest

Processing test of an upgraded mechanical design for PERMCAT reactor

Energy Technology Data Exchange (ETDEWEB)

Borgognoni, Fabio, E-mail: fabio.borgognoni@enea.i [Associazione ENEA-Euratom sulla Fusione, C.R. ENEA Frascati, Via E. Fermi 45, Frascati, Roma I-00044 (Italy); Demange, David; Doerr, Lothar [Forschungszentrum Karlsruhe GmbH, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Postfach 3640, D-76021 Karlsruhe (Germany); Tosti, Silvano [Associazione ENEA-Euratom sulla Fusione, C.R. ENEA Frascati, Via E. Fermi 45, Frascati, Roma I-00044 (Italy); Welte, Stefan [Forschungszentrum Karlsruhe GmbH, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Postfach 3640, D-76021 Karlsruhe (Germany)

2010-12-15

The PERMCAT membrane reactor is a coaxial combination of a Pd/Ag permeator membrane and a catalyst bed. This device has been proposed for processing fusion reactor plasma exhaust gas. A stream containing tritium (up to 1% of tritium in different chemical forms such as water, methane or molecular hydrogen) is decontaminated in the PERMCAT by counter-current isotopic swamping with protium. Different mechanical designs of the membrane reactor have been proposed to improve robustness and lifetime. The ENEA membrane reactor uses a permeator tube with a length of about 500 mm produced via cold-rolling and diffusion welding of Pd/Ag thin foils: two stainless steel pre-tensioned bellows have been applied to the Pd/Ag tube in order to avoid any significant compressive and bending stresses due to the permeator tube elongation consequent to the hydrogen uptake. An experimental test campaign has been performed using this reactor in order to assess the influence of different operating parameters and to evaluate the overall performance (decontamination factor). Tests have been carried out on two reactor prototypes: a defect-free membrane with complete (infinite) hydrogen selectivity and not perm-selective membrane. In this last case, the study has been aimed at verifying the behaviour of the PERMCAT devices under non-normal (accidental) conditions in the view of providing information for future safety analysis. The paper will present the specific mechanical design and the experimental results of tests based on isotopic exchange between H{sub 2}O and D{sub 2}.

Development of three-dimensional pipe bending technology; Pipe zai no sanjigen mage kako gijutsu no kaihatsu

Energy Technology Data Exchange (ETDEWEB)

Takeuchi, K; Takeda, S [Aisin Seiki Co. Ltd., Aichi (Japan)

1997-10-01

Recently, automotive parts uses move resin products or pipe-like products in order to achieve high quality or light weight. Additionally, the shape of automotive parts becomes more complicated. The rotary stretch bending method, although it is most popular method of bending a pipe, has some problems, such as a bending radius is limited due to use of bending mold, a thickness of an outer side of a bending portion is thinner, and a product is scratched easily during manufacturing. We have developed a three dimensional pipe bending process using a floating expanding plug and confirmed that this method can solve the above problems. 2 refs., 9 figs., 3 tabs.

Characterization of the bending strength of craniofacial sutures.

Science.gov (United States)

Maloul, Asmaa; Fialkov, Jeffrey; Whyne, Cari M

2013-03-15

The complex, thin and irregular bones of the human craniofacial skeleton (CFS) are connected together through bony articulations and connective tissues. These articulations are known as sutures and are commonly divided into two groups, facial and cranial sutures, based on their location in the CFS. CFS sutures can exhibit highly variable degrees of interdigitation and complexity and are believed to play a role in accommodating the mechanical demands of the skull. This study aimed to evaluate the mechanical behavior of CFS bone samples with and without sutures and to determine the effect of sutural interdigitations on mechanical strength. Sagittal, coronal, frontozygomatic and zygomaticotemporal sutures along with adjacent bone samples not containing sutures were excised from six fresh-frozen cadaveric heads. The interdigitation of the sutures was quantified through μCT based analysis. Three-point bending to failure was performed on a total of 29 samples. The bending strength of bone samples without sutures demonstrated a non-significant increase of 14% as compared to samples containing sutures (P=0.2). The bending strength of bones containing sutures was positively correlated to the sutural interdigitation index (R=0.701, P=0.002). The higher interdigitation indices found in human cranial vs. facial sutures may be present to resist bending loads as a functional requirement in protecting the brain. Copyright © 2012 Elsevier Ltd. All rights reserved.

[Odontoid bending stiffness after anterior fixation with a single lag screw: biomechanical study].

Science.gov (United States)

Buchvald, P; Čapek, L; Barsa, P

2015-01-01

PURPOSE OF THE STUDY The aim of the experiment was to compare the bending stiffness of an intact odontoid process with bending stiffness after its simulated type II fracture was fixed with a single lag screw. The experiment was done with a desire to answer the question of whether a single osteosynthetic screw is sufficient for good fixation of a type II odontoid fracture. MATERIAL AND METHODS The C2 vertebrae of six cadavers were used. With simultaneous measurement of odontoid bending stiffness, the occurrence of a fracture (type IIA, Grauer's modification of the Anderson- D'Alonzo classification) was simulated using action exerted by a tearing machine in the direction perpendicular to the odontoid axis. Each odontoid fracture was subsequently treated by direct osteosynthesis with a single lag screw inserted in the axial direction by a standard surgical procedure in order to provide conditions similar to those achieved by routine surgical management. The treated odontoid process was subsequently subjected to the same tearing machine loading as applied to it at the start of the experiment. The bending stiffness measured was then compared with that found before the fracture occurred. The results were statistically evaluated by the t-test for paired samples at the level of significance α = 0.05. RESULTS The average value of bending stiffness for odontoid processes of intact vertebrae at the moment of fracture occurrence was 318.3 N/mm. After single axial lag screw fixation of the fracture, the average bending stiffness for the odontoid processes treated was 331.3 N/mm. DISCUSSION Higher values of bending stiffness after screw fixation were found in all specimens and, in comparison with the values recorded before simulated fractures, the increase was statistically significant. CONCLUSIONS The results of our measurements suggest that the single lag screw fixation of a type IIA odontoid fracture will provide better stability for the fracture fragment-C2 body complex on

Mass transfer coefficient factor in pipe bend - 3 D CFD analysis

International Nuclear Information System (INIS)

Prasad, Mahendra; Gaikwad, Avinash J.; Madasamy, P.; Krishnamohan, T.V.; Velumurugan, S.; Sridharan, Arunkumar; Parida, Smrutiranjan

2015-01-01

In power industries Flow Accelerated Corrosion (FAC) has been a concern for pipe wall thinning where high velocity fluid at elevated temperatures is used. Even straight pipes are found to have non uniform corrosion and this is enhanced in junctions such as bends, orifices etc. Mass transfer coefficient (MTC) which defines the amount of corrosion changes from its value in straight pipe (with same fluid parameters) for flow in bends, orifice etc due to changes in velocity profile in axial direction. In this paper, 3 D computational fluid dynamics (CFD) simulation is carried out for an experiment on 58° bend angle and 2D bend radius circular carbon steel pipe carrying water at 120°C under neutral pH conditions. The turbulent model K-ω with shear stress transport was used for this purpose. The mass transfer boundary layer (MTBL) thickness δ mtbl depends on Schmidt number (Sc), as δ mtbl ∼ δ h /(Sc 1/3 ). MTBL is significantly smaller than hydrodynamic boundary layer δ h for large Sc, hence boundary layer meshing was carried out deep into δ mtbl . Uniform velocity was applied at the inlet. The flow velocity was 3 m/s at room temperature while the experimental fluid velocity was 7 m/s. Lower value of fluid velocity is chosen due to the limitations of grid size since it depends inversely on fluid velocity. The ratio of MTC in bend to straight pipe is not strongly dependent on Sc. CFD simulation at lower temperature is sufficient to get approximate MTC in bends. The ratio of the mass transfer coefficient at some locations in bend to the straight pipe coefficient (MTCR) is determined through simulation. The MTC increased in the extrados of the bend towards the outlet. (author)

A preliminary bending fatigue spectrum for steel monostrand cables

DEFF Research Database (Denmark)

Winkler, Jan; Fischer, Gregor; Georgakis, Christos T.

2011-01-01

This paper presents the results of the experimental study on the bending fatigue resistance of high-strength steel monostrand cables. From the conducted fatigue tests in the high-stress, low-cycle region, a preliminary bending fatigue spectrum is derived for the estimation of monostrand cable...... service life expectancy. The presented preliminary bending fatigue spectrum of high-strength monostrands is currently unavailable in the published literature. The presented results provide relevant information on the bending mechanism and fatigue characteristics of monostrand steel cables in tension...... and flexure and show that localized cable bending has a pronounced influence on the fatigue resistance of cables under dynamic excitations....

TPDWR2: thermal power determination for Westinghouse reactors, Version 2. User's guide

International Nuclear Information System (INIS)

Kaczynski, G.M.; Woodruff, R.W.

1985-12-01

TPDWR2 is a computer program which was developed to determine the amount of thermal power generated by any Westinghouse nuclear power plant. From system conditions, TPDWR2 calculates enthalpies of water and steam and the power transferred to or from various components in the reactor coolant system and to or from the chemical and volume control system. From these results and assuming that the reactor core is operating at constant power and is at thermal equilibrium, TPDWR2 calculates the thermal power generated by the reactor core. TPDWR2 runs on the IBM PC and XT computers when IBM Personal Computer DOS, Version 2.00 or 2.10, and IBM Personal Computer Basic, Version D2.00 or D2.10, are stored on the same diskette with TPDWR2

Once-through CANDU reactor models for the ORIGEN2 computer code

International Nuclear Information System (INIS)

Croff, A.G.; Bjerke, M.A.

1980-11-01

Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235 U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given

Radiation protection at the RA Reactor in 1988, Part -2, RA reactor annual report

International Nuclear Information System (INIS)

Ninkovic, M.; Ajdacic, N.; Zaric, M.; Vukovic, Z.

1988-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor and radiation protection; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Decontamination and relevant actions, collecting and treatment of fluid effluents; and and solid radioactive wastes [sr

Foam topology. Bending versus stretching dominated architectures

International Nuclear Information System (INIS)

Deshpande, V.; Ashby, M.; Fleck, N.

2000-01-01

Cellular solids can deform by either the bending or stretching of the cell walls. While most cellular solids are bending-dominated, those that are stretching-dominated are much more weight-efficient for structural applications. In this study we have investigated the topological criteria that dictate the deformation mechanism of a cellular solid by analysing the rigidity (or otherwise) of pin-jointed frameworks comprising inextensional struts. We show that the minimum node connectivity for a special class of lattice structured materials to be stretching-dominated is 6 for 2D foams and 12 for 3D foams. Similarly, sandwich plates comprising of truss cores faced with planar trusses require a minimum node connectivity of 9 to undergo stretching-dominated deformation for all loading states. (author)

Modern Sedimentation off the Kaoping River, SW Taiwan: A Comparison with Eel River's S2S System

Science.gov (United States)

Huh, C.; Lin, H.; Lin, S.

2006-12-01

The Kaoping (KP) River in SW Taiwan has a watershed area of 3257 km2 and an annual sediment discharge of 49 MT. Although the sediment yield of the KP River basin (1.5×104 ton km-2 yr^{- 1}) is the 4th highest among Taiwan's catchment basins, it is nearly one order of magnitude higher than that of the Eel River's basin (~1.8×103 ton km-2 yr-1; the highest in the U.S.). The KP canyon extends almost immediately seaward from the river's mouth and terminates in the northwestern corner of the South China Sea. The head of the canyon is characterized by high and steep walls exceeding 600 m. The KP river's source-to-sink system offers a dramatic case of mountainous rivers at active margins for S2S study. Here we report some results about modern sedimentation in KP river's dispersal system. Seventy-six sediment cores collected from an area of ~3000 km2 were analyzed for fallout nuclides 7Be, 137Cs and 210Pb by gamma spectrometry. From profiles of excess 210Pb and 137Cs sediment accumulation rates in the coring sites were estimated, which vary from 0.06 to 1.6 cm/yr, with the highest rates (>1 cm/yr) distributed in the upper slope (exported out of the study area via the KP canyon to the deep sea by gravity-driven turbidity or hyperpycnal flows.

Effect of Ovality in Inlet Pigtail Pipe Bends Under Combined Internal Pressure and In-Plane Bending for Ni-Fe-Cr B407 Material

Directory of Open Access Journals (Sweden)

Ramaswami P.

2017-09-01

Full Text Available The present paper makes an attempt to depict the effect of ovality in the inlet pigtail pipe bend of a reformer under combined internal pressure and in-plane bending. Finite element analysis (FEA and experiments have been used. An incoloy Ni-Fe-Cr B407 alloy material was considered for study and assumed to be elastic-perfectly plastic in behavior. The design of pipe bend is based on ASME B31.3 standard and during manufacturing process, it is challenging to avoid thickening on the inner radius and thinning on the outer radius of pipe bend. This geometrical shape imperfection is known as ovality and its effect needs investigation which is considered for the study. The finite element analysis (ANSYS-workbench results showed that ovality affects the load carrying capacity of the pipe bend and it was varying with bend factor (h. By data fitting of finite element results, an empirical formula for the limit load of inlet pigtail pipe bend with ovality has been proposed, which is validated by experiments.

Ankle-foot orthosis bending axis influences running mechanics.

Science.gov (United States)

Russell Esposito, Elizabeth; Ranz, Ellyn C; Schmidtbauer, Kelly A; Neptune, Richard R; Wilken, Jason M

2017-07-01

Passive-dynamic ankle-foot orthoses (AFOs) are commonly prescribed to improve locomotion for people with lower limb musculoskeletal weakness. The clinical prescription and design process are typically qualitative and based on observational assessment and experience. Prior work examining the effect of AFO design characteristics generally excludes higher impact activities such as running, providing clinicians and researchers limited information to guide the development of objective prescription guidelines. The proximal location of the bending axis may directly influence energy storage and return and resulting running mechanics. The purpose of this study was to determine if the location of an AFO's bending axis influences running mechanics. Marker and force data were recorded as 12 participants with lower extremity weakness ran overground while wearing a passive-dynamic AFO with posterior struts manufactured with central (middle) and off-centered (high and low) bending axes. Lower extremity joint angles, moments, powers, and ground reaction forces were calculated and compared between limbs and across bending axis conditions. Bending axis produced relatively small but significant changes. Ankle range of motion increased as the bending axis shifted distally (pbenefits during running, although individual preference and physical ability should also be considered. Published by Elsevier B.V.

Bending and tensile deformation of metallic nanowires

International Nuclear Information System (INIS)

McDowell, Matthew T; Leach, Austin M; Gall, Ken

2008-01-01

Using molecular statics simulations and the embedded atom method, a technique for bending silver nanowires and calculating Young's modulus via continuum mechanics has been developed. The measured Young's modulus values extracted from bending simulations were compared with modulus values calculated from uniaxial tension simulations for a range of nanowire sizes, orientations and geometries. Depending on axial orientation, the nanowires exhibit stiffening or softening under tension and bending as size decreases. Bending simulations typically result in a greater variation of Young's modulus values with nanowire size compared with tensile deformation, which indicates a loading-method-dependent size effect on elastic properties at sub-5 nm wire diameters. Since the axial stress is maximized at the lateral surfaces in bending, the loading-method-dependent size effect is postulated to be primarily a result of differences in nanowire surface and core elastic modulus. The divergence of Young's modulus from the bulk modulus in these simulations occurs at sizes below the range in which experiments have demonstrated a size scale effect on elastic properties of metallic nanowires. This difference indicates that other factors beyond native metallic surface properties play a role in experimentally observed nanowire elastic modulus size effects

Irradiation effects on Zr-2.5Nb in power reactors

Energy Technology Data Exchange (ETDEWEB)

Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

Design of a cruciform bend specimen for determination of out-of- plane biaxial tensile stress effects on fracture toughness for shallow cracks

International Nuclear Information System (INIS)

Bass, B.R.; Bryson, J.W.; Mcafee, W.J.; Pennell, W.E.; Theiss, T.J.

1993-01-01

Pressurized-thermal-shock loading in a reactor pressure vessel produces significant positive out-of-plane stresses along the crack front for both circumferential and axial cracks. Experimental evidence, while very limited, seems to indicate that a reduction in toughness is associated with out-of-plane biaxial loading when compared with toughness values obtained under uniaxial conditions. A testing program is described that seeks to determine the effects of out-of-plane biaxial tensile loading on fracture toughness of RPV steels. A cruciform bend specimen that meets specified criteria for the testing pregam is analyzed using three-dimensional elastic-plastic finite-element techniques. These analysis results provide the basis for proposed test conditions that are judged likely to produce a biaxial loading effect in the cruciform bend specimen

River Corridors (Jan 2, 2015)

Data.gov (United States)

Vermont Center for Geographic Information — River corridors are delineated to provide for the least erosive meandering and floodplain geometry toward which a river will evolve over time. River corridor maps...

Modelling the bending/bowing of composite beams such as nuclear fuel

International Nuclear Information System (INIS)

Tayal, M.

1989-01-01

Arrays of tubes are used in many engineered structures, such as in nuclear fuel bundles and in steam generators. The tubes can bend (bow) due to in-service temperatures and loads. Assessments of bowing of nuclear fuel elements can help demonstrate the integrity of fuel and of surrounding components, as a function of operating conditions such as channel power. The BOW code calculates the bending of composite beams such as fuel elements, due to gradients of temperature and due to hydraulic forces. The deflections and rotations are calculated in both lateral directions, for given conditions of temperatures. Wet and dry operation of the sheath can be simulated. BOW accounts for the following physical phenomena: circumferential and axial variations in the temperatures of the sheath and of the pellet; cracking of pellets; grip and slip between the pellets and the sheath; hydraulic drag; restraint from endplates, from neighbouring elements, and from the pressure-tube; gravity; concentric or eccentric welds between endcaps and endplate; neutron flux gradients; and variations of material properties with temperature. The code is based on fundamental principles of mechanics. The governing equations are solved numerically using the finite element method. Several comparisons with closed-form equations shoe that the solutions of BOW are accurate. BOW's predictions for initial in-reactor bow are also consistent with two post-irradiation measurements

IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-07-01

The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1.

IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

International Nuclear Information System (INIS)

1992-01-01

The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1

Bending Strength of EN AC-44200 – Al2O3 Composites at Elevated Temperatures

OpenAIRE

Kurzawa A.; Kaczmar J. W.

2017-01-01

The paper presents results of bend tests at elevated temperatures of aluminium alloy EN AC-44200 (AlSi12) based composite materials reinforced with aluminium oxide particles. The examined materials were manufactured by squeeze casting. Preforms made of Al2O3 particles, with volumetric fraction 10, 20, 30 and 40 vol.% of particles joined with sodium silicate bridges were used as reinforcement. The preforms were characterised by open porosity ensuring proper infiltration with the EN AC-44200 (A...

Smoothed particle hydrodynamics simulations of flow separation at bends

NARCIS (Netherlands)

Hou, Q.; Kruisbrink, A.C.H.; Pearce, F.R.; Tijsseling, A.S.; Yue, T.

2014-01-01

The separated flow in two-dimensional bends is numerically simulated for a right-angled bend with different ratios of the channel widths and for a symmetric bend with different turning angles. Unlike the potential flow solutions that have several restrictive assumptions, the Euler equations are

Smoothed particle hydrodynamics simulations of flow separation at bends

NARCIS (Netherlands)

Hou, Q.; Kruisbrink, A.C.H.; Pearce, F.R.; Tijsseling, A.S.; Yue, T.

2013-01-01

The separated flow in two-dimensional bends is numerically simulated for a right-angled bend with different ratios of the channel widths and for a symmetric bend with different turning angles. Unlike the potential flow solutions that have several restrictive assumptions, the Euler equations are

The bent crystal diffraction spectrometer at the BR2 reactor in Mol

Science.gov (United States)

Kaerts, E.; Jacobs, L.; Vandenput, G.; Van Assche, P. H. M.

1988-05-01

The DuMond-type bent crystal diffraction spectrometer installed at the BR2 reactor in Mol is presented. The spectrometer is mainly designed to study nuclear γ-transitions following thermal neutron capture. It covers the energy interval 25 ≦ Eγ ≦ 1500 keV. Instead of the traditionally used quartz crystals, a highly perfect silicium crystal is chosen as analysing crystal. Diffraction occurs from the (220) plane. The "quasi-mosaic" width, introduced by bending the crystal, is as small as 0.2″. The integrated reflecting power R of the bent crystal stays constant up to 1.5 MeV in first, 680 keV in second and 300 keV in third diffraction order. For higher photon energies, only an E-1 energy dependence is observed in second and third diffraction order. Consequently, besides improving the energy resolution, the use of these silicium crystals substantially increases the spectrometer efficiency and extends the high energy limit of bent crystal diffraction spectrometers. The diffraction angles are measured with a symmetrical interferometer system which covers an angular range of -6° to +6° with a precision of about 0.01″. Minimum diffraction line widths of 0.9″ have been measured, corresponding to an energy resolution ΔE = 1.35 × 10 -6E2n-1 keV -1. The dominant contribution to the observed line widths arises from the finite extent of the source.

The influence of end constraints on smooth pipe bends

International Nuclear Information System (INIS)

Thomson, G.; Spence, J.

1981-01-01

With present trends in the power industries towards higher operating temperatures and pressures, problems associated with the design and safety assessment of pipework systems have become increasingly complex. Within such systems, the importance of smooth pipe bends is well established. The work which will be presented will attempt to clarify the situation and unify the results. An analytical solution of the problem of a linear elastic smooth pipe bend with end constraints under in-plane bending will be presented. The analysis will deal with constraints in the form of flanged tangents of any length. The analysis employs the theorem of minimum total potential energy with suitable kinematically admissible displacements in the form of Fourier series. The integrations and minimisation were performed numerically, thereby permitting the removal of several of the assumptions made by previous authors. Typical results for flexibilities will be given along with comparisons with other works. The differences in some earlier theory are clarified and other more recent work using different solution techniques is substantiated. The bend behaviour is shown to be strongly influenced by the pipe bend parameter, the bend angle, the tangent pipe length and the bend/cross-sectional radius ratio. (orig./GL)

Pilot study risk assessment for selected problems at the Savannah River Site (SRS)

Energy Technology Data Exchange (ETDEWEB)

Hamilton, L.D.; Holtzman, S.; Meinhold, A.; Morris, S.C.; Pardi, R.; Sun, C. [Brookhaven National Lab., Upton, NY (United States); Daniels, J.I.; Layton, D.; McKone, T.E.; Straume, T.; Anspaugh, L. [Lawrence Livermore National Lab., CA (United States)

1993-03-01

An assessment of the health risks was made for releases of tritium and {sup 137}Cs from the Savannah River Site (SRS) at water-receptor locations downriver. Although reactor operations were shut down at the SRS in 1989, liquid wastes continue to be released to the Savannah River either by direct discharges into onsite surface waters or by groundwater transport into surface waters from waste facilities. Existing state mandates will cause the liquid waste streams from future operations to go directly into surface waters. Two drinking water processing plants take water from the river approximately 129 km downriver from the SRS. Potential incremental risks of cancer fatality to individuals and each population were analyzed for either no further reactor operations or resumption of operation of one specific reactor.

Pilot study risk assessment for selected problems at the Savannah River Site (SRS)

Energy Technology Data Exchange (ETDEWEB)

Hamilton, L.D.; Holtzman, S.; Meinhold, A.; Morris, S.C.; Pardi, R.; Sun, C. (Brookhaven National Lab., Upton, NY (United States)); Daniels, J.I.; Layton, D.; McKone, T.E.; Straume, T.; Anspaugh, L. (Lawrence Livermore National Lab., CA (United States))

1993-03-01

An assessment of the health risks was made for releases of tritium and [sup 137]Cs from the Savannah River Site (SRS) at water-receptor locations downriver. Although reactor operations were shut down at the SRS in 1989, liquid wastes continue to be released to the Savannah River either by direct discharges into onsite surface waters or by groundwater transport into surface waters from waste facilities. Existing state mandates will cause the liquid waste streams from future operations to go directly into surface waters. Two drinking water processing plants take water from the river approximately 129 km downriver from the SRS. Potential incremental risks of cancer fatality to individuals and each population were analyzed for either no further reactor operations or resumption of operation of one specific reactor.

The Clinch River Breeder Reactor Plant: an analysis of the impacts of its in-migrant construction workers on local public services. Final report

International Nuclear Information System (INIS)

Braid, R.B. Jr.; Kyles, S.D.

1977-05-01

The socioeconomic impact study identifies certain impacts which are projected to occur to local public services in each of 14 Tennessee communities in the Oak Ridge-Knoxville area during the construction of the Clinch River Breeder Reactor Plant. Various in-migration scenarios are utilized, and detailed qualitative and quantitative analyses of each public service are undertaken. Per capita in-migrant cost-revenue impacts are calculated for each community in each in-migration scenario

Savannah River Laboratory monthly report, November 1991

Energy Technology Data Exchange (ETDEWEB)

Ferrell, J.M. (comp.)

1991-01-01

This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation; tritium facilities and production; separation operations; environmental concerns; and waste management. (FI)

Savannah River Laboratory monthly report, November 1991

Energy Technology Data Exchange (ETDEWEB)

Ferrell, J.M. [comp.

1991-12-31

This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation; tritium facilities and production; separation operations; environmental concerns; and waste management. (FI)

Savannah River Laboratory monthly report, September 1991

Energy Technology Data Exchange (ETDEWEB)

Ferrell, J.M. (comp.)

1991-01-01

This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation, tritium facilities and production; separation operations; environmental concerns; and waste management. (FI)

Savannah River Laboratory monthly report, September 1991

Energy Technology Data Exchange (ETDEWEB)

Ferrell, J.M. [comp.

1991-12-31

This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation, tritium facilities and production; separation operations; environmental concerns; and waste management. (FI)

Savannah River Laboratory monthly report, October 1991

Energy Technology Data Exchange (ETDEWEB)

Ferrell, J.M. (comp.)

1991-01-01

This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation, tritium facilities and production; separations operations; environmental concerns; and waste management. (FI)

Savannah River Laboratory monthly report, October 1991

Energy Technology Data Exchange (ETDEWEB)

Ferrell, J.M. [comp.

1991-12-31

This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation, tritium facilities and production; separations operations; environmental concerns; and waste management. (FI)

OM4 bend insensitive multi-mode fibers’ usefulness for MCM integration

International Nuclear Information System (INIS)

Guzowski, Bartłomiej; Lisik, Zbigniew; Tosik, Grzegorz; Ciupa, Emilia

2012-01-01

Highlights: ► The influence of high temperature exposure on OM4 fibers’ mechanical properties. ► Researching OM4 class fibers for use in innovative Optical Multi Chip Module. ► The influence of bending at a very small radius, up to 2 mm, on MM fibers. - Abstract: For future generations of electronic systems, a severe bottleneck is expected on the interconnection level and the use of optical interconnection is considered as one of the most promising solutions in this matter. Recent progress in fiber development resulted in new generation of optical fibers that are bend insensitive. This makes them ideal for Multi Chip Module (MCM) application. This paper focuses on OM4 bend insensitive multi-mode fibers’ usefulness for MCM integration, particularly the investigation of MM fiber loss is presented, which is influenced by bend diameter and the fiber's mechanical performance under influence of high temperature (400 °C–1000 °C adequate to MCM production process).

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...

Savannah River Technology Center

International Nuclear Information System (INIS)

1993-01-01

This is a monthly progress report from the Savannah River Laboratory for the month of January 1993. It has sections with work in the areas of reactor safety, tritium processes and absorption, separations programs and wastes, environmental concerns and responses, waste management practices, and general concerns

AA, assembly of wide bending magnet

CERN Multimedia

CERN PhotoLab

1980-01-01

The very particular lattice of the AA required 2 types of dipoles (bending magnets; BST, short and wide; BLG, long and narrow). The wide ones had a steel length of 2.71 m, a "good field" width of 0.564 m, and a weight of about 75 t. Here we see the copper coils being hoisted onto the lower half of a BST. See also 7811105, 8006050. For a BLG, see 8001044.

Radionuclide concentrations in white sturgeons from the Hanford Reach of the Columbia River

International Nuclear Information System (INIS)

Dauble, D.D.; Poston, T.M.

1994-01-01

We summarized radionuclide concentrations in white sturgeons Acipenser transmontanus from the Columbia River during a period when several plutonium-production reactors were operating at the Hanford Site in Washington State and compared these values to those measured several years after reactor shutdown. Studies conducted in the Hanford Reach of the Columbia River during 1953-1955 indicated that high concentrations of radionuclides (as total beta) were present in some internal organs on the external surface of white sturgeons. Average concentrations were about 1,480 Bq/kg for liver and kidney and exceeded 2,200 Bq/kg for fins and scutes. The principal radionuclides in the tissues of white sturgeons from the Hanford Reach during 1963-1967, the peak reactor operation interval, were 32 P, 65 Zn, and 51 Cr. Average concentrations of 32 P in muscle ranged from 925 to 2,109 Bq/kg and were typically two to seven times greater than 65 Zn. Average concentrations of radionuclides were usually in the order of gut contents much-gt carcass > muscle. Studies from 1989 to 1990 showed that radionuclide concentrations had decreased dramatically in white sturgeon tissue since the time of reactor operation. Maximum concentrations for artificial radionuclides ( 90 Sr, 60 Co, 137 Cs) in muscle and cartilage of white sturgeons in the Columbia River had declined to less than 4 Bq/kg. Formerly abundant radionuclides, including 32 P, 65 Zn, and 51 Cr, could not be detected in recent tissue samples. Further, radionuclide tissue burden in populations of sturgeons from the Hanford Reach and the upstream or downstream reference locations did not differ significantly. 34 refs., 3 figs., 4 tabs

Minaturized disk bend tests of neutron-irradiated path A type alloys

International Nuclear Information System (INIS)

Lee, M.; Sohn, D.S.; Grant, N.J.; Harling, O.K.

1983-01-01

Path A Prime Candidate Alloy (PCA) has been rapidly solidified and consoliated by extrusion. Twenty percent CW samples, precision TEM disks, 3 phi x 0.254 mm, were irradiated in the mixed flux of the Oak Ridge HFIR reactor up to approx. 8.5 dpa (360 appm He) and approx. 34 dpa (3100 appm He) at 300, 400, 500 and 600 0 C. Similar samples of conventionally processed PCA were also irradiated for comparison. Mechanical properties were characterized using a minaturized disk bend test (MDBT) developed at MIT. These tests indicate major decreases in strength and ductility especially for the 500 and 600 0 C irradiations. No major differences were found between this first version of a rapidly solidified and extruded PCA type alloy and conventionally processed PCA

Mechanical properties and bending strain effect on Cu-Ni sheathed MgB2 superconducting tape

International Nuclear Information System (INIS)

Fu, Minyi; Chen, Jiangxing; Jiao, Zhengkuan; Kumakura, H.; Togano, K.; Ding, Liren; Zhang, Yong; Chen, Zhiyou; Han, Hanmin; Chen, Jinglin

2004-01-01

The Young's modulus (E) of Cu-Ni sheathed MgB 2 monofilament tape was measured using electric method. It is about 8.05 x 10 10 Pa, the same order of Cu and its alloys. We found that the lower E value of the MgB 2 component seemed to relate to the lower filament density. The benefits of pre-compression in filaments were discussed in terms of improving stress distribution in the wires and tapes during winding and operation of superconducting magnets. The magnetic field dependence of J c was investigated on the sample subjected to various strain levels through bending with different radii at 4.2 K

U.S. and foreign breeder reactors

International Nuclear Information System (INIS)

Hill, E.H.

1977-01-01

The running battle between Congress and the Administration over the Clinch River Breeder Reactor Plant (CRBRP) Project has provoked an increased interest in domestic and foreign breeder reactor programs. Perhaps an understanding of the history of breeders here and abroad will serve to place the CRBRP in perspective and allow some analysis of how the U.S. appears on the global canvas. Breeder reactor technology has, for the most part, settled down to concentration on the liquid metal fast breeder reactor (LMFBR). This is the result of 32 years of experience with reactors employing a fast neutron flux and even longer experience with liquid metal coolants. However, a number of U.S. utilities are sponsoring a gas cooled fast reactor program as an alternative technology to the LMFBR. This development program is supported by the U.S. Department of Energy

Characterization of the Three Mile Island Unit-2 reactor building atmosphere prior to the reactor building purge

International Nuclear Information System (INIS)

Hartwell, J.K.; Mandler, J.W.; Duce, S.W.; Motes, B.G.

1981-05-01

The Three Mile Island Unit-2 reactor building atmosphere was sampled prior to the reactor building purge. Samples of the containment atmosphere were obtained using specialized sampling equipment installed through penetration R-626 at the 358-foot (109-meter) level of the TMI-2 reactor building. The samples were subsequently analyzed for radionuclide concentration and for gaseous molecular components (O 2 , N 2 , etc.) by two independent laboratories at the Idaho National Engineering Laboratory (INEL). The sampling procedures, analysis methods, and results are summarized

FFTF and CRBRP reactor vessels

International Nuclear Information System (INIS)

Morgan, R.E.

1977-01-01

The Fast Flux Test Facility (FFTF) reactor vessel and the Clinch River Breeder Reactor Plant (CRBRP) reactor vessel each serve to enclose a fast spectrum reactor core, contain the sodium coolant, and provide support and positioning for the closure head and internal structure. Each vessel is located in its reactor cavity and is protected by a guard vessel which would ensure continued decay heat removal capability should a major system leak develop. Although the two plants have significantly different thermal power ratings, 400 megawatts for FFTF and 975 megawatts for CRBRP, the two reactor vessels are comparable in size, the CRBRP vessel being approximately 28% longer than the FFTF vessel. The FFTF vessel diameter was controlled by the space required for the three individual In-Vessel Handling Machines and Instrument Trees. Utilization of the triple rotating plug scheme for CRBRP refueling enables packaging of the larger CRBRP core in a vessel the same diameter as the FFTF vessel

Radiation protection at the RA Reactor in 1998, RA reactor annual report, Part -2

International Nuclear Information System (INIS)

Ninkovic, M.; Pavlovic, R.; Mandic, M.; Pavlovic, S.; Grsic, Z.

1998-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Collecting and treatment of fluid effluents; and (4) radioactive wastes, decontamination and actions. Each of the category is described as a separate annex of this report [sr

Models of bending strength for Gilsocarbon graphites irradiated in inert and oxidising environments

International Nuclear Information System (INIS)

Eason, Ernest D.; Hall, Graham N.; Marsden, Barry J.; Heys, Graham B.

2013-01-01

This paper presents the development and validation of an empirical model of fast neutron damage and radiolytic oxidation effects on bending strength for the moulded Gilsocarbon graphites used in Advanced Gas-cooled Reactors (AGRs). The inert environment model is based on evidence of essentially constant strength as fast neutron dose increases in inert environment. The model of combined irradiation and oxidation calibrates that constant along with an exponential function representing the degree of radiolytic oxidation as measured by weight loss. The change in strength with exposure was found to vary from one AGR station to another. The model was calibrated to data on material trepanned from AGR moderator bricks after varying operating times

Bends and splitters in graphene nanoribbon waveguides

DEFF Research Database (Denmark)

Zhu, Xiaolong; Yan, Wei; Mortensen, N. Asger

2013-01-01

We investigate the performance of bends and splitters in graphene nanoribbon waveguides. Although the graphene waveguides are lossy themselves, we show that bends and splitters do not induce any additional loss provided that the nanoribbon width is sub-wavelength. We use transmission line theory...

Control rod supporting device in reactor

International Nuclear Information System (INIS)

Seki, Osamu; Itooka, Satoshi; Harada, Kiyoshi; Jodoi, Takashi.

1990-01-01

Since coolants flowing from a reactor core hit against a control rod and a control rod connection pipe, a considerable amount of bending moment for separating an attracting surface between an electromagnet and an armature is formed. Then, a plurality of grooves are formed on a heat sensitive material to dispose a heat collecting fin, and each of upper and lower contact portions of a control rod supporting portion in which the flanged portion of T-like cross section does not slip out is made into a partial spheric surface and a portion between the electromagnet and the attracted member are engaged by the unevenness. With such a constitution, even if a bending moment is applied, the control rod only swings and the bending moment is not transmitted to the attracted member. Further, since the temperature of the heat sensitive material can be rapidly made closer to the peripheral temperature by using the heat collecting fin, the timing for separation is made accurate. Further, since the engaging portion is brought into contact at the spheric surface, the load distribution on the control rod is made uniform, and the positional relationship is made accurate, to support the control rod reliably and the separation depends only on the temperature of the coolants. (N.H.)

Two case studies in river naturalization: planform migration and bank erosion control

Science.gov (United States)

Abad, J. D.; Guneralp, I.; Rhoads, B. L.; Garcia, M. H.

2005-05-01

A sound understanding of river planform evolution and bank erosion control, along with integration of expertise from several disciplines is required for the development of predictive models for river naturalization. Over the last few years, several methodologies have been presented for naturalization projects, from purely heuristic to more advanced methods. Since the time and space scales of concern in naturalization vary widely, there is a need for appropriate tools at a variety of time and space scales. This study presents two case studies at different scales. The first case study describes the prediction of river planform evolution for a remeandering project based on a simplified two-dimensional hydrodynamic model. The second case study describes the applicability of a Computational Fluid Dynamics (CFD) model for evaluating the effectiveness of bank-erosion control structures in individual meander bends. Understanding the hydrodynamic influence of control structures on flow through bends allows accurate prediction of depositional and erosional distribution patterns, resulting in better assessment on river planform stability, especially for the case of natural complex systems. The first case study introduces a mathematical model for evolution of meandering rivers that can be used in remeandering projects. In United States in particular, several rivers have been channelized in the past causing environmental and ecological problems. Following Newton's third law, "for every action, there is a reaction", naturalization techniques evolve as natural reactive solutions to channelization. This model (herein referred as RVR Meander) can be used as a stand-alone Windows application or as module in a Geographic Information System. The model was applied to the Poplar Creek re-meanderization project and used to evaluate re-meandering alternatives for an approximately 800-meter long reach of Poplar Creek that was straightened in 1938. The second case study describes a

Active species in a large volume N2-O2 post-discharge reactor

International Nuclear Information System (INIS)

Kutasi, K; Pintassilgo, C D; Loureiro, J; Coelho, P J

2007-01-01

A large volume post-discharge reactor placed downstream from a flowing N 2 -O 2 microwave discharge is modelled using a three-dimensional hydrodynamic model. The density distributions of the most populated active species present in the reactor-O( 3 P), O 2 (a 1 Δ g ), O 2 (b 1 Σ g + ), NO(X 2 Π), NO(A 2 Σ + ), NO(B 2 Π), NO 2 (X), O 3 , O 2 (X 3 Σ g - ) and N( 4 S)-are calculated and the main source and loss processes for each species are identified for two discharge conditions: (i) p = 2 Torr, f = 2450 MHz, and (ii) p = 8 Torr, f = 915 MHz; in the case of a N 2 -2%O 2 mixture composition and gas flow rate of 2 x 10 3 sccm. The modification of the species relative densities by changing the oxygen percentage in the initial gas mixture composition, in the 0.2%-5% range, are presented. The possible tuning of the species concentrations in the reactor by changing the size of the connecting afterglow tube between the active discharge and the large post-discharge reactor is investigated as well

Mesohabitats, fish assemblage composition, and mesohabitat use of the Rio Grande silvery minnow over a range of seasonal flow regimes in the Rio Grande/Rio Bravo del Norte, in and near Big Bend National Park, Texas, 2010-11

Science.gov (United States)

Moring, J. Bruce; Braun, Christopher L.; Pearson, Daniel K.

2014-01-01

In 2010–11, the U.S. Geological Survey (USGS), in cooperation with the U.S. Fish and Wildlife Service, evaluated the physical characteristics and fish assemblage composition of mapped river mesohabitats at four sites on the Rio Grande/Rio Bravo del Norte (hereinafter Rio Grande) in and near Big Bend National Park, Texas. The four sites used for the river habitat study were colocated with sites where the U.S. Fish and Wildlife Service has implemented an experimental reintroduction of the Rio Grande silvery minnow (Hybognathus amarus), a federally listed endangered species, into part of the historical range of this species. The four sites from upstream to downstream are USGS station 08374340 Rio Grande at Contrabando Canyon near Lajitas, Tex. (hereinafter the Contrabando site), USGS station 290956103363600 Rio Grande at Santa Elena Canyon, Big Bend National Park, Tex. (hereinafter the Santa Elena site), USGS station 291046102573900 Rio Grande near Ranger Station at Rio Grande Village, Tex. (hereinafter the Rio Grande Village site), and USGS station 292354102491100 Rio Grande above Stillwell Crossing near Big Bend National Park, Tex. (hereinafter the Stillwell Crossing site).

G 2 reactor project; Projet de pile a double fin: G 2

Energy Technology Data Exchange (ETDEWEB)

Ailleret, [Electricite de France (EDF), Dir. General des Etudes de Recherches, 75 - Paris (France); Taranger, P; Yvon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The CEA actually constructs the G-2 reactor core working with natural uranium, which will use graphite as moderator, and gas under pressure as cooling fluid. This report presents the specificity of the new reactor: - the different elements of the reactor core, - the control and the security of the reactor, - the renewal of the fuel, - the biologic surrounding wall, - and the cooling circuit. (M.B.) [French] le Commissariat a l'Energie Atomique construit actuellement la pile G-2 a Uranium naturel, qui utilisera le graphite comme moderateur, et le gaz sous pression comme fluide de refroidissement. Ce rapport presente les specificite du nouveau reacteur: - les differents elements de la pile, - le controle et la securite du reacteur, - le renouvellement du combustible, - l'enceinte biologique, - et le circuit de refroidissement. (M.B.)

Reactor handbook. 2. rev. ed.

International Nuclear Information System (INIS)

Lederer, B.J.; Wildberg, D.W.

1992-01-01

On the basis of the guidelines on expert knowledge, the book discusses the subjects of atomic physics, heat transfer, nuclear power plants, reactor materials, radiation protection, reactor safety, reactor instrumentation, and reactor operation, with special regard to nuclear power plants with LWR-type reactors. The book is intended for shift personnel, especially gang bosses, reactor operators, and control station operators: for this reason a practical and rather popular style has been chosen. However, the book will also be a manual for other operating personnel, personnel of producer companies, expert organisations, authorities, and students. It can be used as a textbook for staff training, a manual for the practice, and as accompanying book for teaching at nuclear engineering schools. (orig.) With 173 figs [de

Environment-friendly reduction of flood risk and infrastructure damage in a mountain river: Case study of the Czarny Dunajec

Science.gov (United States)

Mikuś, Paweł; Wyżga, Bartłomiej; Radecki-Pawlik, Artur; Zawiejska, Joanna; Amirowicz, Antoni; Oglęcki, Paweł

2016-11-01

Migration of a mountain river channel may cause erosional risk to infrastructure or settlements on the valley floor. Following a flood of 2010, a cutbank in one of the bends of the main channel of the Czarny Dunajec, Polish Carpathians, approached a local road by 50 m. To arrest the erosion of the laterally migrating channel, water authorities planned construction of a ditch cutting the forested neck of the bend, reinforcement of the ditch banks, and damming the main channel with a boulder groyne. In order to avoid channelization of the highly valued, multithread river reach that would deteriorate its ecological status and cause increased flood risk to downstream reaches, an alternative approach to prevent bank erosion was proposed. The new scheme, applied in 2011, included opening of the inlets to inactive side braids located by the neck of the bend of the main channel. This solution reestablished the flow in the steeper low-flow channels, allowing us to expect a cutoff and abandonment of the main channel during subsequent floods. Gravelly deflectors were constructed directly below the inlets to the reactivated side channels to divert the flow into the channels and prevent the water from entering the main channel. Hydraulic measurements performed before and after the implementation of the scheme confirmed that it enabled shifting the main water current, with the highest average velocity and bed shear stress, from the braid closest to the road to the most distant braid. Similar surveys of fish and benthic macroinvertebrate communities indicated that flow reactivation in the side channels was beneficial for these groups of river biota, increasing their abundance and taxonomic richness in the reach. Not only was the implemented solution significantly less expensive, but it also enhanced ecological functions of the multithread channel and the variability of physical habitat conditions and maintained the role of the reach as a wood debris trap. However, avulsion of the

Thermal effects on the Savannah River

International Nuclear Information System (INIS)

Patrick, R.

1981-01-01

The effects of thermal effluents from the Savannah River Plant (SRP), particularly during periods when the L Reactor was operative, on the structure and health of the aquatic communities of organisms in the Savannah River have been determined. Portions of the data base collected by the Academy of Natural Sciences since 1951 on the Savannah River were used. The organisms belonging to various groups of aquatic life were identified to species if possible. The relative abundance of the species was estimated for the more common species. The bacteriological, chemical and physical characteristics of the water were determined

Channel modelling and performance analysis of V2I communication systems in blind bend scattering environments

KAUST Repository

Chelli, Ali

2014-01-01

In this paper, we derive a new geometrical blind bend scattering model for vehicle-to- infrastructure (V2I) communications. The proposed model takes into account single-bounce and double- bounce scattering stemming from fixed scatterers located on both sides of a curved street. Starting from the geometrical blind bend model, the exact expression of the angle of departure (AOD) is derived. Based on this expression, the probability density function (PDF) of the AOD and the Doppler power spectrum are determined. Analytical expressions for the channel gain and the temporal autocorrelation function (ACF) are provided under non-line-of-sight (NLOS) conditions. Additionally, we investigate the impact of the position of transmitting vehicle relatively to the receiving road-side unit on the channel statistics. Moreover, we study the performance of different digital modulations over a sum of singly and doubly scattered (SSDS) channel. Note that the proposed V2I channel model falls under the umbrella of SSDS channels since the transmitted signal undergoes a combination of single-bounce and double-bounce scattering. We study some characteristic quantities of SSDS channels and derive expressions for the average symbol error probability of several modulation schemes over SSDS channels with and without diversity combining. The validity of these analytical expressions is confirmed by computer-based simulations.

Safety analysis for K reactor and impact of cooling tower installation

International Nuclear Information System (INIS)

Fields, C.C.; Wooten, L.A.; Geeting, M.W.; Morgan, C.E.; Buczek, J.A.; Smith, D.C.

1993-01-01

This paper describes the safety analysis of the Savannah River site K-reactor loss-of-cooling-water-supply (LOCWS) event and the impact on the analysis of a natural-draft cooling tower, which was installed in 1992. Historically (1954 to 1992), the K-reactor secondary cooling system [called the cooling water system (CWS)] used water from the Savannah River pumped to a 25-million-gal basin adjacent to the reactor. Approximately 170 000 gal/min were pumped from the basin through heat exchangers to remove heat from the primary cooling system. This water then entered a smaller basin, where it flowed over a weir and eventually returned to the Savannah River. The 25-million-gal basin is at a higher elevation than the heat exchangers and the smaller basin to supply cooling by gravity flow (which is sufficient to remove decay heat) if power to the CWS pumps is interrupted. Small amounts of cooling water are also used for other essential equipment such as diesels, motors, and oil coolers. With the cooling tower installed, ∼85% of the cooling water flows from the small basin by gravity to the cooling tower instead of returning to the Savannah River. After being cooled, it is pumped back to the 25-million-gal basin. River water is supplied only to make up for evaporation and the blowdown stream

Social support modifies association between forward bending of the trunk and low-back pain

DEFF Research Database (Denmark)

Villumsen, Morten; Holtermann, Andreas; Samani, Afshin

2016-01-01

OBJECTIVES: This study aimed to investigate the association between forward bending of the trunk and low-back pain intensity (LBPi) among blue-collar workers in Denmark as well as whether the level of social support modifies the association. METHODS: In total, 457 workers were included in the study...... support was categorized into low, moderate, and high levels. Multi-adjusted logistic regressions estimated the association between forward bending and LBPi and the effect modification by social support. RESULTS: Forward bending and LBPi were not significantly associated but modified by social support....... Workers with low social support and long duration of forward bending had higher likelihood of high LBPi [odds ratio (OR) 2.97, 95% confidence interval (95% CI) 1.11-7.95] compared to workers with high social support and long duration of forward bending. Among workers with low social support, workers...

Preconstruction radioactivity levels in the vicinity of the proposed Clinch River Breeder Reactor Project

International Nuclear Information System (INIS)

1984-05-01

Routine samples of ground water, river water, and bottom sediment were collected from the Clinch River in 1983 in the preconstruction-construction phase of the CRBRP environmental radiological monitoring program. The water samples analyzed for iodine-131 yielded only a slight indication of the presence of I-131 at levels below the nominal lower limit of detection of 0.5 pCi/L. The only significant radioisotopes identified in sediment samples were 137 Cs, 60 Co, and the naturally occurring 40 K. The results for 137 Cs vary from 2.2 to 10.1 pCi/g (dry weight), while the results for 60 Co range from 0.35 to 1.2 pCi/g (dry weight). With the exception of tritium, no significant radioactivity was detected in ground or surface water at the CRBRP site. Tritium concentrations ranging from 12,667 to 12,823 pCi/L were found in samples of surface water taken from the Clinch River below Melton Hill Dam while samples taken at the dam exhibited tritium levels from 28 to 942 pCi/L. These elevated tritium levels in the Clinch River below Melton Hill Dam are attributable to DOE operations at Oak Ridge. The external gamma radiation levels measured at the CRBRP site averaged 17.4 +- 3.2 mR/quarter for 1983. This is consistent with levels measured at TVA's nonoperating nuclear power plant construction sites. 3 figures, 8 tables

Characterization and study of photonic crystal fibres with bends

International Nuclear Information System (INIS)

Belhadj, W.; AbdelMalek, F.; Bouchriha, H.

2006-01-01

Analysis of a photonic crystal fibre (PRCF) with bends is presented. Using the versatile finite difference time domain method, the modal characteristics of the PCFs are found. Possibilities of employing PCFs with bends in sensing are discussed. It is found that a large evanescent field is present when the bend angle exceeds 45 o

Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

Energy Technology Data Exchange (ETDEWEB)

Montes Ponce de Leon, J.

1974-07-01

In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

Piezoelectric micromotor based on the structure of serial bending arms.

Science.gov (United States)

Tong, Jianhua; Cui, Tianhong; Shao, Peige; Wang, Liding

2003-09-01

This paper presents a new piezoelectric micromotor based on the structure of serial bending arms. Serial bending arms are composed of two piezoelectric bimorphs with one end fixed and the other end free, driven by two signals of a biased square wave with a phase difference of pi/2. The free end of a cantilever arm will move along an elliptic orbit so that the cantilever is used to drive a cylinder rotor. The rotor's end surface contacts the free end of the cantilever, resulting in the rotor's rotation. There are six serial bending arms anchored on the base. The driving mechanism of the micromotor is proposed and analyzed. A new micromotor prototype, 5 mm in diameter, has been fabricated and characterized. The maximum rotational speed reaches 325 rpm, and the output torque is about 36.5 microNm.

Development of and verification test integral reactor major components - Development of manufacturing process and fabrication of prototype for SG and CEDM

Energy Technology Data Exchange (ETDEWEB)

Lee, Chang Hee; Park, Hwa Kyu; Kim, Yong Kyu; Choi, Yong Soon; Kang, Ki Su; Hyun, Young Min [Korea Heavy Industries and Construction Co., LTD., Changwon (Korea)

1999-03-01

Integral SMART(System integrated Modular Advanced Reactor) type reactor is under conceptual design. Because major components is integrated within in a single pressure vessel, compact design using advanced technology is essential. It means that manufacturing process for these components is more complex and difficult. The objective of this study is to confirm the possibility of manufacture of Steam Generator, Control Element Drive Mechanism(CEDM) and Reactor Assembly which includes Reactor Pressure Vessel, it is important to understand the design requirement and function of the major components. After understanding the design requirement and function, it is concluded that the helical bending and weld qualification of titanium tube for Steam Generator and the applicability of electron beam weld for CEDM step motor parts is the critical to fabricate the components. Therefore, bending mock-up and weld qualification of titanium tube was performed and the results are quite satisfactory. Also, it is concluded that electron beam welding technique can be applicable to the CEDM step motor part. (author). 22 refs., 14 figs., 46 tabs.

Tensile and bending fatigue of the adhesive interface to dentin.

Science.gov (United States)

Belli, Renan; Baratieri, Luiz Narciso; Braem, Marc; Petschelt, Anselm; Lohbauer, Ulrich

2010-12-01

The aim of this study was to evaluate the fatigue limits of the dentin-composite interfaces established either with an etch-and-rinse or an one-step self-etch adhesive systems under tensile and bending configurations. Flat specimens (1.2 mm×5 mm×35 mm) were prepared using a plexiglass mold where dentin sections from human third molars were bonded to a resin composite, exhibiting the interface centrally located. Syntac Classic and G-Bond were used as adhesives and applied according to the manufacturer's instructions. The fluorochrome Rhodamine B was added to the adhesives to allow for fractographic evaluation. Tensile strength was measured in an universal testing machine and the bending strength (n=15) in a Flex machine (Flex, University of Antwerp, Belgium), respectively. Tensile (TFL) and bending fatigue limits (BFL) (n=25) were determined under wet conditions for 10(4) cycles following a staircase approach. Interface morphology and fracture mechanisms were observed using light, confocal laser scanning and scanning electron microscopy. Statistical analysis was performed using three-way ANOVA (mod LSD test, pTensile and bending characteristic strengths at 63.2% failure probability for Syntac were 23.8 MPa and 71.5 MPa, and 24.7 MPa and 72.3 MPa for G-Bond, respectively. Regarding the applied methods, no significant differences were detected between adhesives. However, fatigue limits for G-Bond (TFL=5.9 MPa; BFL=36.2 MPa) were significantly reduced when compared to Syntac (TFL=12.6 MPa; BFL=49.7 MPa). Fracture modes of Syntac were generally of adhesive nature, between the adhesive resin and dentin, while G-Bond showed fracture planes involving the adhesive-dentin interface and the adhesive resin. Cyclic loading under tensile and bending configurations led to a significant strength degradation, with a more pronounced fatigue limit decrease for G-Bond. The greater decrease in fracture strength was observed in the tensile configuration. Copyright © 2010 Academy of

Study to evaluate the feasibility of constructing a retrofit containment for the 105-L reactor at the Savannah River Plant

International Nuclear Information System (INIS)

Quinn, R.D.

1989-01-01

This paper presents a summary of a study performed to determine the feasibility of constructing a retrofit containment dome meeting the requirements of the ASME Boiler and Pressure Vessel Code for nuclear containment vessels over the existing Savannah River 105-L reactor. Using existing large dome structures as a guide, design concepts were developed and analyses performed to evaluate the structural feasibility of containment dome structures. Construction schedules and costs were estimated to assess financial feasibility as well. It was concluded that such a retrofit containment dome was structurally feasible and within the capabilities of present day construction technology

Small specimen test technology of fracture toughness in structural material F82H steel for fusion nuclear reactors

International Nuclear Information System (INIS)

Wakai, Eiichi; Ohtsuka, Hideo; Jitsukawa, Shiro; Matsukawa, Shingo; Ando, Masami

2006-03-01

Small specimen test technology (SSTT) has been developed to investigate mechanical properties of nuclear materials. SSTT has been driven by limited availability of effective irradiation volumes in test reactors and accelerator-based neutron and charged particle sources, and it is very useful for the reduction of waste materials produced in nuclear engineering. In this study new bend test machines have been developed to obtain fracture behaviors of F82H steel for very small bend specimens of pre-cracked t/2-1/3CVN (Charpy V-notch) with 20 mm-length and DFMB (deformation and fracture mini bend specimen) with 9 mm-length and disk compact tension of 0.18DCT type, and fracture behaviors were examined to evaluate DBTT (ductile-brittle transition temperature) at temperature from -180 to 25degC. The effect of specimen size on DBTT of F82H steel was also examined by using Charpy type specimens such as 1/2t-CVN, 1/3CVN and t/2-1/3CVN. In this paper, it also provides the information of the specimens irradiated at 250degC and 350degC to about 2 dpa in the capsule of 04M-67A and 04M-68A of JMTR experiments. (author)

Interactions between Point Bar Growth and Bank Erosion on a Low Sinuosity Meander Bend in an Ephemeral Channel: Insights from Repeat Topographic Surveys and Numerical Modeling

Science.gov (United States)

Ursic, M.; Langendoen, E. J.

2017-12-01

Interactions between point bar growth, bank migration, and hydraulics on meandering rivers are complicated and not well understood. For ephemeral streams, rapid fluctuations in flow further complicate studying and understanding these interactions. This study seeks to answer the following `cause-and-effect' question: Does point bar morphologic adjustment determine where bank erosion occurs (for example, through topographic steering of the flow), or does local bank retreat determine where accretion/erosion occurs on the point bar, or do bank erosion and point bar morphologic adjustment co-evolve? Further, is there a response time between the `cause-and-effect' processes and what variables determine its magnitude and duration? In an effort to answer these questions for an ephemeral stream, a dataset of forty-eight repeat topographic surveys over a ten-year period (1996-2006) of a low sinuosity bend within the Goodwin Creek Experimental Watershed, located near Batesville, MS, were utilized in conjunction with continuous discharge measurements to correlate flow variability and erosional and depositional zones, spatially and temporally. Hydraulically, the bend is located immediately downstream of a confluence with a major tributary. Supercritical flumes on both the primary and tributary channels just upstream of the confluence provide continuous measured discharges to the bend over the survey period. In addition, water surface elevations were continuously measured at the upstream and downstream ends of the bend. No spatial correlation trends could be discerned between reach-scale bank retreat, point bar morphologic adjustment, and flow discharge. Because detailed flow patterns were not available, the two-dimensional computer model Telemac2D was used to provide these details. The model was calibrated and validated for a set of runoff events for which more detailed flow data were available. Telemac2D simulations were created for each topographic survey period. Flows

The distribution characteristics of pollutants released at different cross-sectional positions of a river

International Nuclear Information System (INIS)

Huang Heqing; Chen Guang; Zhang Qianfeng

2010-01-01

The distribution characteristics of heavier or lighter pollutants released at different cross-sectional positions of a wide river is investigated with a well-tested three-dimensional numerical model of gravity flows based on Reynolds-Averaged Navier-Stokes equations and turbulence k-ε model. By focusing on investigating the influences of flow and buoyancy on pollutants, it is found that while carrying by the river flow downstream: i) a heavier pollutant released from the cross-sectional side position, forms transverse oscillation between two banks with decreased amplitude, i.e. forms kind of helical flow pattern along the straight part of channel bed; ii) a heavier pollutant released from the cross-sectional middle position, forms collapse oscillation in the middle of the straight channel part with reduced amplitude; iii) in the downstream sinuous channel, heavier pollutant is of higher concentration on the outer side of channel bends; iv) a light pollutant released from the cross-sectional side position, slips partly to the other side of the river, resulting in higher concentrations on two sides of the channel top; v) a light pollutant released from the cross-sectional middle position, splits into two parts symmetrically along two sides of the channel top; vi) in the downstream sinuous channel, light pollutant presents higher concentration on the inner side of channel bends. These findings may assist in cost-effective scientific countermeasures to be taken for accidental or planned pollutant releases into a river. - The distribution characteristics of heavier or lighter pollutants released at different cross-sectional positions of a river.

Disk-bend ductility tests for irradiated materials

International Nuclear Information System (INIS)

Klueh, R.L.; Braski, D.N.

1984-01-01

We modified the HEDL disk-bend test machine and are using it to qualitatively screen alloys that are susceptible to embrittlement caused by irradiation. Tests designed to understand the disk-bend test in relation to a uniaxial test are discussed. Selected results of tests of neutron-irradiated material are also presented

Structural analyses on piping systems of sodium reactors. 2. Eigenvalue analyses of hot-leg pipelines of large scale sodium reactors

International Nuclear Information System (INIS)

Furuhashi, Ichiro; Kasahara, Naoto

2002-01-01

Two types of finite element models analyzed eigenvalues of hot-leg pipelines of a large-scale sodium reactor. One is a beam element model, which is usual for pipe analyses. The other is a shell element model to evaluate particular modes in thin pipes with large diameters. Summary of analysis results: (1) A beam element model and a order natural frequency. A beam element model is available to get the first order vibration mode. (2) The maximum difference ratio of beam mode natural frequencies was 14% between a beam element model with no shear deformations and a shell element model. However, its difference becomes very small, when shear deformations are considered in beam element. (3) In the first order horizontal mode, the Y-piece acts like a pendulum, and the elbow acts like the hinge. The natural frequency is strongly affected by the bending and shear rigidities of the outer supporting pipe. (4) In the first order vertical mode, the vertical sections of the outer and inner pipes moves in the axial-directional piston mode, the horizontal section of inner pipe behaves like the cantilever, and the elbow acts like the hinge. The natural frequency is strongly affected by the axial rigidity of outer supporting pipe. (5) Both effective masses and participation factors were small for particular shell modes. (author)