Inherent safety that the reactivity effect of core bending in fast reactors brings about

International Nuclear Information System (INIS)

Nakagawa, Masatoshi; Yagawa, Genki.

1994-01-01

FBRs have the merit on safety by low operation pressure and the large heat capacity of coolant, in addition, due to the core temperature rise at the time of accidents and the thermal expansion of core structures, the negative feedback of reactivity can be expected. Recently, attention has been paid to the negative feedback of reactivity due to core bending. It can be expected also in the core of limited free bow type. Bending is caused by the difference of thermal expansion on six surfaces of hexagonal wrapper tubes. The bending changes core reactivity and exerts effects to fuel exchange force and operation, insertion of control rods and the structural soundness of fuel assemblies. for the purpose of limiting the effect that core bending exerts to core characteristics to allowable range, core constraint mechanism is installed. The behavior of core bending at the time of anticipated transient without scram is explained. The example of the analysis of PRISM reactor is shown. The experiment that confirmed the negative feedback of reactivity due to core bending under the condition of ULOF was that at the fast flux test facility. (K.I.)

Applicability of ANSYS ELBOW290 element for flexibility calculation of tight radius bends on feeder pipes in CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Zhang, X., E-mail: Xuan.Zhang@candu.com [Candu Energy Inc, Mississauga, ON (Canada)

2015-07-01

A curved pipe element, ELBOW290, became available in ANSYS 12. This element was developed based on a simplified shell theory, and maintains the ability to capture cross-sectional deformations of elbows. Numerical testing on the applicability of this element for the flexibility calculation of the tight radius bends in CANDU reactors is carried out to determine the usability of this element in completing stress analyses for feeder pipes. Comparisons are made between the ELBOW290 and the shell element for various feeder bend types found in domestic and overseas CANDU reactors. The comparisons show that the ELBOW290 element is suitable for calculating the flexibility of the tight radius bends. (author)

A Study on U-bending Technology using Rotary Draw Bending

Energy Technology Data Exchange (ETDEWEB)

Kwak, Ok-gyu; Kim, Won-seok [BHI Co., Gyunsang-Namdo (Korea, Republic of); Ku, Tae-wan [Pusan National Univ., Busan (Korea, Republic of)

2014-10-15

In the steam generator, heat transfer phenomenon for producing the steam between the primary system of the nuclear reactor and the secondary one occurs around the heat transfer tube. That is, the primary coolant with high temperature(320 .deg.. C) and high pressure(157Kgf/cm2) derived from the reactor flows in the heat transfer tube, and the secondary one runs out that tube. Therefore, it is able to mention that the heat transfer tube itself is a boundary of the heat transfer phenomenon. The heat transfer tube bundle of each steam generator used for the PWR and the PHWR(Pressurized Heavy Water Reactor) is generally composed of about 8,000-13,000 U-tubes. And these tubes are the core component as the structural and heat transfer material in the steam generator, which is in charge of cooling about 70% of the cooling surface of the primary system. For achieving the U-bending process with the thin walled tube, generally, a mandrel could be inserted in the tube according to the bending radius. But when the bending radius is small, the tube U-bending process could be also performed without the mandrel. In this study, numerical and experimental investigations on the U-bending process for producing the heat transfer tubes by using the straight and long tubes were carried out with the consideration of the elastic recovery after the U-bending. In the numerical approach, finite element analysis scheme was adopted with a commercial code, ABAQUS Implicit/Explicit. As the precedent study, the related experiment was also performed to verify the predicted results on the ovality and the minimum wall thickness of the U-bending heat transfer tube. Furthermore, its bending process was also conducted to analyze the deformation behavior for the Alloy 690 tube. In this study, the U-bending process was considered to simulate and manufactured the heat transfer tube used for the steam generator. To investigate the deformation behavior of the U-bending process, and a series of the

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

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

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.

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.

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

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)

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

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

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

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.)

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

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

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)

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.)

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

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

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.)

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

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

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.

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

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

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

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.)

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

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...

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

Optimization of superconducting bending magnets for a 1.0 to 1.5 GeV compact light source

International Nuclear Information System (INIS)

Green, M.A.; Garren, A.A.

1995-06-01

Compact light sources are being proposed for protein crystallography, medical imaging, nano-machining and other areas of study that require intense sources of x rays at energies up to 35 keV. In order for a synchrotron light source to be attractive, its capital cost must, be kept low. The proposed compact light source has superconducting bending elements to bend the stored beam and produce the x rays. Additional focusing for the machine is provided by conventional quadrupoles. An important part of the cost optimization of a compact light source is the cost of the bending magnets. In the case of a machine with superconducting bending elements, the bending magnet system can represent close to half of the storage ring cost. The compact light source storage rings studied here have a range of stored electron energies from 1.0 to 1.5 GeV. For a number of reasons, it is desirable to keep the storage ring circumference below 30 meters. Cost optimization parameters include: (1) the number of superconducting bending elements in the ring, and (2) the central induction of the dipole. A machine design that features two superconducting dipoles in a single cryostat vacuum vessel is also discussed

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

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

Missouri River Flood 2011 Vulnerabilities Assessment Report. Volume 2 - Technical Report

Science.gov (United States)

2012-10-01

Michels at Dakota Dunes , South Dakota. ............................................................................................................... 2...91 Figure 28. Upper Hamburg Bend Levee Toe Scour...Bend Project at Dakota Dunes along Left Bank River Mile 737 ........................... 109 Figure 37. Stage Trends on the Missouri River at St

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

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)

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

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

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

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

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

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

Thirtieth anniversary of reactor accident in A-1 Nuclear Power Plant Jaslovske Bohunice

International Nuclear Information System (INIS)

Kuruc, J.; Matel, L.

2007-01-01

The facts about reactor accidents in A-1 Nuclear Power Plant Jaslovske Bohunice, Slovakia are presented. There was the reactor KS150 (HWGCR) cooled with carbon dioxide and moderated with heavy water. A-1 NPP was commissioned on December 25, 1972. The first reactor accident happened on January 5, 1976 during fuel loading. This accident has not been evaluated according to the INES scale up to the present time. The second serious accident in A-1 NPP occurred on February 22, 1977 also during fuel loading. This INES level 4 of reactor accident resulted in damaged fuel integrity with extensive corrosion damage of fuel cladding and release of radioactivity into the plant area. The A-1 NPP was consecutively shut down and is being decommissioned in the present time. Both reactor accidents are described briefly. Some radioecological and radiobiological consequences of accidents and contamination of area of A-1 NPP as well as of Manivier Canal and Dudvah River as result of flooding during the decommissioning are presented (authors)

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

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

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

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.

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

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.

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

Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors (U)

International Nuclear Information System (INIS)

Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

1990-01-01

This paper reports on a full-scope probabilistic risk assessment (PRA) performed for the Savannah River Site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident

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

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

Study on reactor building structure using ultrahigh strength materials, 1

International Nuclear Information System (INIS)

Ishimura, Kikuo; Odajima, Masahiro; Irino, Kazuo; Hashiba, Toshio.

1991-01-01

This study was promoted to be aimed at realization of the optimal nuclear reactor building structure of the future. As the first step, the study regarding ultrahigh strength reinforced concrete (abbr. RC) shear wall was selected. As the result of various tests, the application of ultrahigh strength RC shear walls was verified. The tests conducted were relevant to; ultrahigh strength concrete material tests; pure shear tests of RC flat panels; and bending shear tests and its simulation analysis of RC shear walls. (author)

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.

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.)

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.

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)

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

TRAC-PF1/MOD3 calculations of Savannah River Laboratory Rig FA single-annulus heated experiments

International Nuclear Information System (INIS)

Fischer, S.R.; McDaniel, C.K.

1992-01-01

This paper presents the results of TRAC-PF1/MOD3 benchmarks of the Rig FA experiments performed at the Savannah River Laboratory to simulate prototypic reactor fuel assembly behavior over a range of fluid conditions typical of the emergency cooling system (ECS) phase of a loss-of-coolant accident (LOCA). The primary purpose of this work was to use the SRL Rig FA tests to qualify the TRAC-PF1/MOD3 computer code and models for computing Mark-22 fuel assembly LOCA/ECS power limits. This qualification effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to independently confirm power limits for the Savannah River Site K Reactor. The results of this benchmark effort as discussed in this paper demonstrate that TRAC/PF1/MOD3 coupled with proper modeling is capable of simulating thermal-hydraulic phenomena typical of that encountered in Mark-22 fuel assembly during LOCA/ECS conditions

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....

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

Magnetic design considerations for the SSC vertical bending (BV1C) magnet

International Nuclear Information System (INIS)

Venkatraman, V.; Goodzeit, C.; Jayakumar, R.; Nobrega, F.; Snitchler, G.

1994-01-01

The BV1C magnet is a large aperture, vertical bending magnet to be used to bend proton beams in the interaction region. An aperture larger than 80 mm is required. The central field has to be a minimum of 6T with a 10% margin. The lattice requirements for field quality are stringent because two counter beams traverse this magnet off the center axis. This magnet's transfer function sag is specified to match closely the transfer function sag of the low beta quadrupoles. With these specifications in mind, suitable designs for the 2-D magnetic cross-sections have been analyzed

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

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

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

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

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

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

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

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

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.

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

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

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...

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

Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors

International Nuclear Information System (INIS)

Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

1990-01-01

A full-scope probabilistic risk assessment (PRA) is being performed for the Savannah River site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident. The SRS PRA has three principal objectives: improved understanding of SRS reactor safety issues through discovery and understanding of the mechanisms involved. Improved risk management capability through tools for assessing the safety impact of both current standard operations and proposed revisions. A quantitative measure of the risks posed by SRS reactor operation to employees and the general public, to allow comparison with declared goals and other societal risks

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

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,

Tests for development of estimation technology of reactor core deformation. Report No.1: fundamental mechanical properties of wrapper tube (test report)

International Nuclear Information System (INIS)

Nishiura, Takeo; Shimazaki, Yuji; Horikiri, Morito

1998-10-01

Mechanical properties such as local contact compression stiffness, bending stiffness, deformation properties, material properties, and friction properties of a wrapper tube structure were clarified experimentally, which can be used as the basic data for development of estimation technology of reactor core deformation. Contents of the Tests data as follows: (1) Effects of load supporting boundary conditions, whether or not a contact-proof pad is attached, and length of duct, on cross section deformation of wrapper tube were made clear as the local contact compression stiffness characteristics. (2) Bending stiffness does not depend on the difference of load supporting boundary conditions. The property of cross section deformation under bending load was obtained. (3) The deformation modes and the strain distributions were obtained by the deformation tests of wrapper tube. (4) The stress-strain diagrams including plastic range under various strain variation rates were obtained by the material tests at room temperature. (5) The static and the dynamic friction coefficients by various contact angles and the contact loads between contact-proof pads of two wrapper tubes were obtained by friction property tests. (author)

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

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

Project management plan for the 105-C Reactor interim safe storage project. Revision 1

International Nuclear Information System (INIS)

Miller, R.L.

1997-01-01

In 1942, the Hanford Site was commissioned by the US Government to produce plutonium. Between 1942 and 1955, eight water-cooled, graphite-moderated reactors were constructed along the Columbia River at the Hanford Site to support the production of plutonium. The reactors were deactivated from 1964 to 1971 and declared surplus. The Surplus Production Reactor Decommissioning Project (BHI 1994b) will decommission these reactors and has selected the 105-C Reactor to be used as a demonstration project for interim safe storage at the present location and final disposition of the entire reactor core in the 200 West Area. This project will result in lower costs, accelerated schedules, reduced worker exposure, and provide direct benefit to the US Department of Energy for decommissioning projects complex wide. This project sets forth plans, organizational responsibilities, control systems, and procedures to manage the execution of the Project Management Plan for the 105-C Reactor Interim Safe Storage Project (Project Management Plan) activities to meet programmatic requirements within authorized funding and approved schedules. The Project Management Plan is organized following the guidelines provided by US Department of Energy Order 4700.1, Project Management System and the Richland Environmental Restoration Project Plan (DOE-RL 1992b)

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.

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

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.

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.

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

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

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

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.

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

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.

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.

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.

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

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

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.

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

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

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

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

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

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.

Training courses at VR-1 reactor

International Nuclear Information System (INIS)

Sklenka, L.; Kropik, M.

2006-01-01

This paper describes one of the main purposes of the VR-1 training reactor utilization - i.e. extensive educational program. The educational program is intended for the training of university students and selected nuclear power plant personnel. The training courses provide them experience in reactor and neutron physics, dosimetry, nuclear safety and operation of nuclear facilities. At present, the training course participants can go through more than 20 standard experimental exercises; particular exercises for special training can be prepared. Approximately 200 university students become familiar with the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. The VR-1 reactor takes also part in Eugene Wigner Course on Reactor Physics Experiments in the framework of European Nuclear Educational Network (ENEN) association. Recently, training courses for Bulgarian research reactor specialists supported by IAEA were carried out. An attractive program including demonstration of reactor operation is prepared also for high school students. Every year, more than 1500 high school students come to visit the reactor, as do many foreigner visitors. (author)

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.

Current status of the Thai Research Reactor (TRR-1/M1)

International Nuclear Information System (INIS)

Chueinta, Siripone; Julanan, Mongkol; Charncanchee, Decharchai

2006-01-01

The first Thai Research Reactor, TRR-1 went critical on 27 October 1962 at the maximum power of 1 MW. It was located at Office of Atoms for Peace (OAP) in Bangkok. Since then, TRR-1 was continuously operated and eventually shut down in 1975. Plate type, high-enriched uranium (HEU) and U 3 O 8 A1 cladding were used as the reactor fuel. Light water was used as moderator and coolant as well. In 1975, because of the problem from fuel supplier and also to supporting the Treaty of Non Proliferation of Nuclear Weapon or NPT, TRR-1 was shut down for modification. The reactor core and control system were disassembled and replaced by TRIGA Mark III. A new core was a hexagonal core shape designed by General Atomics (GA). Afterwards, TRR-1 was officially renamed to the Thai Research Reactor-1/Modification 1 (TRR-1/M1). TRR-1/M1 is a multipurpose swimming pool type reactor with nominal power of 2 MW. The TRR-1/M1 uses uranium enriched at 20% in U-235 (LEU) and ZrH alloy as fuel. Light water is also used as coolant and moderator. At present, the reactor is operating with core No.14. The reactor has been serving for various kinds of utilization namely, radioisotope production, neutron activation analysis, beam experiments and reactor physics experiments. (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

Fourier Transform Infrared spectrum of the OCD bending mode in methanol-D1

Science.gov (United States)

Mukhopadhyay, Indra

2016-03-01

The infrared (IR) spectra corresponding to OCD bending vibration of asymmetrically deuterated methanol species CH2DOH have been recorded with a Fourier Transform Spectrometer. The spectrum shows a typical structure of a parallel a-type band. This is expected because the bending vibration mainly executed parallel to the symmetry axis The Q-branch lines are grouped closely around 896 cm-1 and the P- and R-Branches show complex structure. Nonetheless it was possible to assign a-type P- and R-branch lines up to K value of 8 and J value up to about 20 in most cases. The Q-branch lines for higher K values can be followed to about J = 15, the presence of which confirmed the assignments. The observations suggest that in the OCD bend some energy levels are highly interacted by highly excited torsional state from the ground torsional state. A full catalogue is presented along with the effective molecular parameters. An intensity anomaly was also observed in the transitions. So far it has been possible to assign only transitions between e0 ← e0 states. Plausible explanations of intensity anomaly are presented. Lastly, a number of optically pumped far infrared (FIR) laser lines have been assigned either to exact or tentative quantum states. These assignments should prove valuable for production of new FIR laser lines.

Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

Energy Technology Data Exchange (ETDEWEB)

Montes, J

1972-07-01

In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

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%

Rate estimates for lateral bedrock erosion based on radiocarbon ages, Duck River, Tennessee

International Nuclear Information System (INIS)

Brakenridge, G.R.

1985-01-01

Rates of bedrock erosion in ingrown meandering rivers can be inferred from the location of buried relict flood-plain and river-bank surfaces, associated paleosols, and radiocarbon dates. Two independent methods are used to evaluate the long-term rates of limestone bedrock erosion by the Duck River. Radiocarbon dates on samples retrieved from buried Holocene flood-plain and bank surfaces indicate lateral migration of the river bank at average rates of 0.6-1.9 m/100 yr. Such rates agree with lateral bedrock cliff erosion rates of 0.5-1.4 m/100 yr, as determined from a comparison of late Pleistocene and modern bedrock cliff and terrace scarp positions. These results show that lateral bedrock erosion by this river could have occurred coevally with flood-plain and terrace formation and that the resulting evolution of valley meander bends carved into bedrock is similar in many respects to that of channel meanders cut into alluvium. 11 references, 5 figures

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.)

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)

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

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.

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.

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

Annual report on JEN-1 reactor

International Nuclear Information System (INIS)

Montes, J.

1972-01-01

In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

Structural integrity of water reactor pressure boundary components. Progress report ending 29 February 1976

International Nuclear Information System (INIS)

Loss, F.J.

1976-01-01

The report describes progress in the following areas: (a) fatigue crack propagation in reactor pressure vessel steels in an air environment, (b) dynamic fracture toughness of 1-in. (25-mm) and precracked Charpy-V bend specimens under impact loading, (c) postirradiation notch ductility and properties recovery in reactor vessel steels, (d) factors contributing to variable resistance of structural steels to radiation embrittlement, and (e) the initial program plan to investigate the phenomena of warm prestress and plastic net ligament in support of thermal shock studies

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

Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Kastenberg, W.E.; Solomon, K.A.

1979-07-01

In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

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. 

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

Safety operation of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.

2001-01-01

There are three nuclear research reactors in the Czech Republic in operation now: light water reactor LVR-15, maximum reactor power 10 MW t , owner and operator Nuclear Research Institute Rez; light water zero power reactor LR-0, maximum reactor power 5 kW t , owner and operator Nuclear Research Institute Rez and training reactor VR-1 Sparrow, maximum reactor power 5 kW t , owner and operate Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. The training reactor VR-1 Vrabec 'Sparrow', operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly it is designed for training the students of Czech universities, preparing the experts for the Czech nuclear programme, as well as for certain research work, and for information programmes in the nuclear programme, as well as for certain research work, and for information programmes in sphere of using the nuclear energy (public relations). (author)

Savannah River reactor process water heat exchanger tube structural integrity margin Task Number 92-005-1

International Nuclear Information System (INIS)

Mertz, G.E.; Barnes, D.M.; Sindelar, R.L.

1992-02-01

Twelve process water heat exchangers are designed to remove heat generated in the reactor tank. Each heat exchanger has approximately 9000, 1/2 inch diameter x 0.049 inches thick tubes. Minimum structural tubing requirements and the leak rate through postulated tubing defects are developed in this report A comparison of the structural requirements and the defect size calculated to produce leak rates of 0.5 lbs./day demonstrate adequate structural margins against gross tube rupture. Commercial nuclear experience with pressurized water reactor (PWR) steam generator plugging criteria are used for guidance in performing this analysis. It is important to note that the SRS reactors are low energy systems with normal operating pressures of 203 psig at 130 degree F while the PWR is a high energy system with operating pressures near 2200 psig at 600 degree F. Clearly the PVM steam generator has loadings which are more severe than the SRS heat exchangers. Consistent with the Regulatory Guide 1.121 criteria both wastage (wall thinning) and cracking are addressed. Structural limits on wall thinning and crack size are developed to preclude gross rupture. ASME Section XI criteria, with the factors of safety recommended by Regulatory Guide 1.121 are used to develop the allowable crack size criteria. Normal operating conditions (pressure, dead weight, and hydraulic drag) are considered with seismic and water hammer accident conditions. Both the wall thinning and crack size criteria are developed for the end-of-evaluation period. Allowances for corrosion, wear, or crack growth have not been included in this analysis Structurally, the tubing is over designed and can tolerate large defects with adequate margins against gross rupture. The structural margins of heat exchanger tubing are evident by contrasting the tubing's structural capacity, per the ASME Code, with its operating conditions/configuration

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.

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

Extensive utilization of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, Karel; Sklenka, Lubomir

2003-01-01

Full text: The training reactor VR-1 Vrabec ('Sparrow'), operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly, it is designed and operated for training of students from Czech universities, preparing of experts for the Czech nuclear programme, as well as for certain research and development work, and for information programmes in the sphere of non-military nuclear energy use (public relation). The VR-1 training reactor is a pool-type light-water reactor based on enriched uranium with maximum thermal power 1kWth and short time period up to 5kW th . The moderator of neutrons is light demineralized water (H 2 O) that is also used as a reflector, a biological shielding, and a coolant. Heat is removed from the core with natural convection. The reactor core contains 14 to 18 fuel assemblies IRT-3M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The core is accommodated in a cylindrical stainless steel vessel - pool, which is filled with water. UR-70 control rods serve the reactor control and safe shutdown. Training of the VR-1 reactor provides students with experience in reactor and neutron physics, dosimetry, nuclear safety, and nuclear installation operation. Students from technical universities and from natural sciences universities come to the reactor for training. Approximately 200 university students are introduced to the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. Practical Course on Reactor Physics in Framework of European Nuclear Engineering Network has been newly introduced. Currently, students can try out more than 20 experimental exercises. Further training courses have been included

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

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)

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

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

Magnetic field-induced elastic bending in bilayers of Tb1−xDyxFe2−y and Pb(Zr1−zTiz)O3

International Nuclear Information System (INIS)

Jin, Tao; Qichao, Wu; Ning, Zhang

2014-01-01

Magnetic field-induced strain in the magnetoelectric bilayers of Tb 1−x Dy x Fe 2−y and Pb(Zr 1−z Ti z )O 3 was studied. A butterfly shaped strain curve was observed on the surface of Pb(Zr 1−z Ti z )O 3 . The shape of the strain curve was found to be related to the sample thickness and the volume fraction occupied by the ferroelectrics in the bilayer. Theoretical analysis and experimental results showed that magnetoelastic bending in the bilayer composites was largely responsible for the butterfly strain curve. - Highlights: • Butterfly strain curves were observed on the PZT surface for bilayers of TDF and PZT. • The strain curve is related to the sample thickness and the volume fraction of the PZT. • A physics model depicting the field-controlled bending of the bilayers was developed. • The magnetoelastic bending was found to account for the butterfly strain curve

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

Assessment of plutonium in the Savannah River Site environment. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Carlton, W.H.; Evans, A.G.; Geary, L.A.; Murphy, C.E. Jr.; Pinder, J.E.; Strom, R.N.

1992-12-31

Plutonium in the Savannah River Site Environment is published as a part of the Radiological Assessment Program (RAP). It is the fifth in a series of eight documents on individual radioisotopes released to the environment as a result of Savannah River Site (SRS) operations. These are living documents, each to be revised and updated on a two-year schedule. This document describes the sources of plutonium in the environment, its release from SRS, environmental transport and ecological concentration of plutonium, and the radiological impact of SRS releases to the environment. Plutonium exists in the environment as a result of above-ground nuclear weapons tests, the Chernobyl accident, the destruction of satellite SNAP 9-A, plane crashes involving nuclear weapons, and small releases from reactors and reprocessing plants. Plutonium has been produced at SRS during the operation of five production reactors and released in small quantities during the processing of fuel and targets in chemical separations facilities. Approximately 0.6 Ci of plutonium was released into streams and about 12 Ci was released to seepage basins, where it was tightly bound by clay in the soil. A smaller quantity, about 3.8 Ci, was released to the atmosphere. Virtually all releases have occurred in F- and H-Area separation facilities. Plutonium concentration and transport mechanisms for the atmosphere, surface water, and ground water releases have been extensively studied by Savannah River Technology Center (SRTC) and ecological mechanisms have been studied by Savannah River Ecology Laboratory (SREL). The overall radiological impact of SRS releases to the offsite maximum individual can be characterized by a total dose of 15 mrem (atmospheric) and 0.18 mrem (liquid), compared with the dose of 12,960 mrem from non-SRS sources during the same period of time (1954--1989). Plutonium releases from SRS facilities have resulted in a negligible impact to the environment and the population it supports.

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.

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).

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.)

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

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

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.

Moving ring reactor 'Karin-1'

International Nuclear Information System (INIS)

1983-12-01

The conceptual design of a moving ring reactor ''Karin-1'' has been carried out to advance fusion system design, to clarify the research and development problems, and to decide their priority. In order to attain these objectives, a D-T reactor with tritium breeding blanket is designed, a commercial reactor with net power output of 500 MWe is designed, the compatibility of plasma physics with fusion engineering is demonstrated, and some other guideline is indicated. A moving ring reactor is composed mainly of three parts. In the first formation section, a plasma ring is formed and heated up to ignition temperature. The plasma ring of compact torus is transported from the formation section through the next burning section to generate fusion power. Then the plasma ring moves into the last recovery section, and the energy and particles of the plasma ring are recovered. The outline of a moving ring reactor ''Karin-1'' is described. As a candidate material for the first wall, SiC was adopted to reduce the MHD effect and to minimize the interaction with neutrons and charged particles. The thin metal lining was applied to the SiC surface to solve the problem of the compatibility with lithium blanket. Plasma physics, the engineering aspect and the items of research and development are described. (Kako, I.)

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

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

The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

Energy Technology Data Exchange (ETDEWEB)

Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

2008-07-15

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (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

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.

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

Safety of research reactors (Design and Operation)

International Nuclear Information System (INIS)

Dirar, H. M.

2012-06-01

The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

Measurements with the Chalk River Calorimeters

International Nuclear Information System (INIS)

Boyd, A.W.

1970-01-01

The Chalk River calorimeters were designed to measure the absorbed dose rate in reactors in materials such as graphite, polyethylene and beryllium in the range 0.01-1 Wg -1 . To eliminate heaters in the sample they were made to operate adiabatically, or more accurately quasi-adiabatically since there is no heater on the jacket. Both the sample and jacket temperatures are recorded from the time of insertion in the reactor flux and the absorbed dose rate is calculated from these data. The advantages of this type of calorimeter are the ease of construction and the absence of a sample heater. The disadvantage is that dose rates below ~ 10 mWg -1 cannot be determined accurately

Occupational analysis for the Angra-1 reactor

International Nuclear Information System (INIS)

Moraes, A.

1991-01-01

Due to several modifications which were imposed to its time schedule during construction, the Angra-1 reactor did not enter to the grid in 1982 as it was initially foreseen. These modifications occurred due to an unforeseen scenario that was verified in steam generators (serie D-3, Westinghouse) of power stations with similar configurations which had been installed in other countries such as Ringhals-3 (Sweden), Almaraz-1 (Spain) and McGuine-1 (USA). So, among the main events that occurred in the Angra-1 reactor, which were of interest from the point of view of radiation protection, it could be pointed out the personnel monitoring, and the occupational exposure measurements at different reactor power, during the reactor fueling and during modification and tests performed at the steam generators and at ducts of the primary coolant circuit. (author)

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

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

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

Reactor operations at SAFARI-1

International Nuclear Information System (INIS)

Vlok, J.W.H.

2003-01-01

A vigorous commercial programme of isotope production and other radiation services has been followed by the SAFARI-1 research reactor over the past ten years - superimposed on the original purpose of the reactor to provide a basic tool for nuclear research, development and education to the country at an institutional level. A combination of the binding nature of the resulting contractual obligations and tighter regulatory control has demanded an equally vigorous programme of upgrading, replacement and renovation of many systems in order to improve the safety and reliability of the reactor. Not least among these changes is the more effective training and deployment of operations personnel that has been necessitated as the operational demands on the reactor evolved from five days per week to twenty four hours per day, seven days per week, with more than 300 days per year at full power. This paper briefly sketches the operational history of SAFARI-1 and then focuses on the training and structuring currently in place to meet the operational needs. There is a detailed step-by-step look at the operator?s career plan and pre-defined milestones. Shift work, especially the shift cycle, has a negative influence on the operator's career path development, especially due to his unavailability for training. Methods utilised to minimise this influence are presented. The increase of responsibilities regarding the operation of the reactor, ancillaries and experimental facilities as the operator progresses with his career are discussed. (author)

Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Owen, M.B.

1997-04-01

This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

International Nuclear Information System (INIS)

Owen, M.B.

1997-04-01

This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future

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

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).

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.)

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

Galvanic vestibular stimulation may improve anterior bending posture in Parkinson's disease.

Science.gov (United States)

Okada, Yohei; Kita, Yorihiro; Nakamura, Junji; Kataoka, Hiroshi; Kiriyama, Takao; Ueno, Satoshi; Hiyamizu, Makoto; Morioka, Shu; Shomoto, Koji

2015-05-06

This study investigated the effects of binaural monopolar galvanic vestibular stimulation (GVS), which likely stimulates the bilateral vestibular system, on the anterior bending angle in patients with Parkinson's disease (PD) with anterior bending posture in a single-blind, randomized sham-controlled crossover trial. The seven PD patients completed two types of stimulation (binaural monopolar GVS and sham stimulation) applied in a random order 1 week apart. We measured each patient's anterior bending angles while he or she stood with eyes open and eyes closed before/after the stimulations. The anterior bending angles in both the eyes-open and the eyes-closed conditions were significantly reduced after the GVS. The amount of change in the eyes-closed condition post-GVS was significantly larger than that by sham stimulation. The amount of change in anterior bending angles in the GVS condition was not significantly correlated with Unified Parkinson's Disease Rating Scale motor score, disease duration, the duration of the postural deformities, and the anterior bending angles before the GVS. Binaural monopolar GVS might improve anterior bending posture in PD patients, irrespective of the duration and the severity of disease and postural deformities. Binaural monopolar GVS might be a novel treatment strategy to improve anterior bending posture in PD.

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

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)

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

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.

Extensive utilization of training reactor VR-1

International Nuclear Information System (INIS)

Karel, Matejka; Lubomir, Sklenka

2005-01-01

This paper describes one of the main purposes of the VR-1 training reactor utilisation - i.e. extensive educational programme. The educational programme is intended for the training of university students (all technical universities in Czech Republic) and selected nuclear power plant personnel. At the present, students can go through more than 20 different experimental exercises. An attractive programme including demonstration of reactor operation is prepared also for high school students. Moreover, research and development works and information programmes proceed at the VR-1 reactor as well

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

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

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.

A Temperature Sensor Based on a Polymer Optical Fiber Macro-Bend

Science.gov (United States)

Moraleda, Alberto Tapetado; García, Carmen Vázquez; Zaballa, Joseba Zubia; Arrue, Jon

2013-01-01

The design and development of a plastic optical fiber (POF) macrobend temperature sensor is presented. The sensor has a linear response versus temperature at a fixed bend radius, with a sensitivity of 1.92·10−3 (°C)−1. The sensor system used a dummy fiber-optic sensor for reference purposes having a resolution below 0.3 °C. A comprehensive experimental analysis was carried out to provide insight into the effect of different surrounding media on practical macro-bend POF sensor implementation. Experimental results are successfully compared with bend loss calculations. PMID:24077323

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

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

Effect of Bend Radius on Magnitude and Location of Erosion in S-Bend

Directory of Open Access Journals (Sweden)

Quamrul H. Mazumder

2015-01-01

Full Text Available Solid particle erosion is a mechanical process that removes material by the impact of solid particles entrained in the flow. Erosion is a leading cause of failure of oil and gas pipelines and fittings in fluid handling industries. Different approaches have been used to control or minimize damage caused by erosion in particulated gas-solid or liquid-solid flows. S-bend geometry is widely used in different fluid handling equipment that may be susceptible to erosion damage. The results of a computational fluid dynamic (CFD simulation of diluted gas-solid and liquid-solid flows in an S-bend are presented in this paper. In addition to particle impact velocity, the bend radius may have significant influence on the magnitude and the location of erosion. CFD analysis was performed at three different air velocities (15.24 m/s–45.72 m/s and three different water velocities (0.1 m/s–10 m/s with entrained solid particles. The particle sizes used in the analysis range between 50 and 300 microns. Maximum erosion was observed in water with 10 m/s, 250-micron particle size, and a ratio of 3.5. The location of maximum erosion was observed in water with 10 m/s, 300-micron particle size, and a ratio of 3.5. Comparison of CFD results with available literature data showed reasonable and good agreement.

Stability analysis of the Ghana Research Reactor-1 (GHARR-1)

International Nuclear Information System (INIS)

Della, R.; Alhassan, E.; Adoo, N.A.; Bansah, C.Y.; Nyarko, B.J.B.; Akaho, E.H.K.

2013-01-01

Highlights: • We developed a theoretical model to study the stability of the Ghana Research Reactor-1. • The neutronics transfer function was described by the point kinetics model for a single group of delayed neutrons. • The thermal hydraulics transfer function was based on the modified lumped parameter concept. • A computer code, RESA (REactor Stability Analysis) was developed. • Results show that the closed-loop transfer function was stable and well damped for variable operating power levels. - Abstract: A theoretical model has been developed to study the stability of the Ghana Research Reactor one (GHARR-1). The closed-loop transfer function of GHARR-1 was established based on the model, which involved the neutronics and the thermal hydraulics transfer functions. The reactor kinetics was described by the point kinetics model for a single group of delayed neutrons, whilst the thermal hydraulics transfer function was based on the modified lumped parameter concept. The inherent internal feedback effect due to the fuel and the coolant was represented by the fuel temperature coefficient and the moderator temperature coefficient respectively. A computer code, RESA (REactor Stability Analysis), entirely in Java was developed based on the model for systems analysis. Stability analysis of the open-loop transfer function of GHARR-1 based on the Nyquist criterion and Bode diagrams using RESA, has shown that the closed-loop transfer function was marginally stable for variable operating power levels. The relative stability margins of GHARR-1 were also identified

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.

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

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

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.

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

Numerical modeling of a nuclear production reactor cooling lake

International Nuclear Information System (INIS)

Hamm, L.L.; Pepper, D.W.

1987-01-01

A finite element model has been developed which predicts flow and temperature distributions within a nuclear reactor cooling lake at the Savannah River Plant near Aiken, South Carolina. Numerical results agree with values obtained from a 3-D EPA numerical lake model and actual measurements obtained from the lake. Because the effluent water from the reactor heat exchangers discharges directly into the lake, downstream temperatures at mid-lake could exceed the South Carolina DHEC guidelines for thermal exchanges during the summer months. Therefore, reactor power was reduced to maintain temperature compliance at mid-lake. Thermal mitigation measures were studied that included placing a 6.1 m deep fabric curtain across mid-lake and moving the reactor outfall upstream. These measurements were calculated to permit about an 8% improvement in reactor power during summer operation

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...

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)

Estimation of radioactivity in structural materials of ETRR-1 reactor

Energy Technology Data Exchange (ETDEWEB)

Imam, M [National Center for Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

1997-12-31

Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig.

Polluted Alamuyo River: Impacts on surrounding wells, microbial ...

African Journals Online (AJOL)

SERVER

2008-02-19

Feb 19, 2008 ... (A) and down stream (F) of the river were also evaluated using ... pathogens isolated from the wastewater can survive in ... and control were fixed in ethanol- glacial acetic acid (3:1, v/v). The ..... aquatic toixity and characterization of chemical and micro- ... wastewater in anaerobic sequencing batch reactors.

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.

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

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.

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.

Effect of cross section on collapse load in pipe bends subjected to in ...

African Journals Online (AJOL)

user

Also various researchers have estimated the plastic loads of pipe bends with cracks (Hong et al, ... In reality, the pipe bend exists with shape imperfections namely ovality and ... C t. −. = ×. (3). 3. Finite element limit analysis. Figure 1. Pipe bend with ..... Chattopadhyay J., Natahani D. K., Dutta B. K. and Kushwaha H. S. 2000.

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

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.'

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.

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)

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.

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...

Colloid-colloid hydrodynamic interaction around a bend in a quasi-one-dimensional channel.

Science.gov (United States)

Liepold, Christopher; Zarcone, Ryan; Heumann, Tibor; Rice, Stuart A; Lin, Binhua

2017-07-01

We report a study of how a bend in a quasi-one-dimensional (q1D) channel containing a colloid suspension at equilibrium that exhibits single-file particle motion affects the hydrodynamic coupling between colloid particles. We observe both structural and dynamical responses as the bend angle becomes more acute. The structural response is an increasing depletion of particles in the vicinity of the bend and an increase in the nearest-neighbor separation in the pair correlation function for particles on opposite sides of the bend. The dynamical response monitored by the change in the self-diffusion [D_{11}(x)] and coupling [D_{12}(x)] terms of the pair diffusion tensor reveals that the pair separation dependence of D_{12} mimics that of the pair correlation function just as in a straight q1D channel. We show that the observed behavior is a consequence of the boundary conditions imposed on the q1D channel: both the single-file motion and the hydrodynamic flow must follow the channel around the bend.

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)

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.

Tritium in surface water of the Yenisei river Basin

International Nuclear Information System (INIS)

Bondareva, L.G.; Bolsunovsky, A.Ya.

2005-01-01

The paper reports an investigation of the tritium content in the surface waters of the Yenisei River basin near the Mining-and-Chemical Combine (MCC). In 2001-2003 the maximum tritium concentration in the Yenisei River did not exceed 4±1 Bq/L. It has been found that there are surface waters containing enhanced tritium, up to 168 Bq/L, as compared with the background values for the Yenisei River. There are two possible sources of tritium input. First, the last operating reactor of the MCC, which still uses the Yenisei water as coolant. Second, tritium may come from the deep aquifers at the Severny testing site. For the first time tritium has been found in two aquatic plant species of the Yenisei River with maximal tritium concentration 304 Bq/Kg wet weight. Concentration factors of tritium for aquatic plants are much higher than 1

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...

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...

Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

International Nuclear Information System (INIS)

Lee, D.D.; Collins, J.L.

2000-01-01

One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required

Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

Energy Technology Data Exchange (ETDEWEB)

Lee, D.D.; Collins, J.L.

2000-02-01

One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.

Operation characteristics and conditions of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.; Kolros, A.; Polach, S.; Sklenka, L.

1994-01-01

The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

Bend Faulting at the Edge of a Flat Slab: The 2017 Mw7.1 Puebla-Morelos, Mexico Earthquake

Science.gov (United States)

Melgar, Diego; Pérez-Campos, Xyoli; Ramirez-Guzman, Leonardo; Spica, Zack; Espíndola, Victor Hugo; Hammond, William C.; Cabral-Cano, Enrique

2018-03-01

We present results of a slip model from joint inversion of strong motion and static Global Positioning System data for the Mw7.1 Puebla-Morelos earthquake. We find that the earthquake nucleates at the bottom of the oceanic crust or within the oceanic mantle with most of the moment release occurring within the oceanic mantle. Given its location at the edge of the flat slab, the earthquake is likely the result of bending stresses occurring at the transition from flat slab subduction to steeply dipping subduction. The event strikes obliquely to the slab, we find a good agreement between the seafloor fabric offshore the source region and the strike of the earthquake. We argue that the event likely reactivated a fault first created during seafloor formation. We hypothesize that large bending-related events at the edge of the flat slab are more likely in areas of low misalignment between the seafloor fabric and the slab strike where reactivation of preexisting structures is favored. This hypothesis predicts decreased likelihood of bending-related events northwest of the 2017 source region but also suggests that they should be more likely southeast of the 2017 source region.

Tunable characteristics of bending resonance frequency in magnetoelectric laminated composites

Institute of Scientific and Technical Information of China (English)

Chen Lei; Li Ping; Wen Yu-Mei; Zhu Yong

2013-01-01

As the magnetoelectric (ME) effect in piezoelectric/magnetostrictive laminated composites is mediated by mechanical deformation,the ME effect is significantly enhanced in the vicinity of resonance frequency.The bending resonance frequency (fr) of bilayered Terfenol-D/PZT (MP) laminated composites is studied,and our analysis predicts that (i) the bending resonance frequency of an MP laminated composite can be tuned by an applied dc magnetic bias (Hdc) due to the △E effect; (ii) the bending resonance frequency of the MP laminated composite can be controlled by incorporating FeCuNbSiB layers with different thicknesses.The experimental results show that with Hdc increasing from 0Oe (1 Oe=79.5775 A/m)to 700 Oe,the bending resonance frequency can be shifted in a range of 32.68 kHz ≤ fr ≤ 33.96 kHz.In addition,with the thickness of the FeCuNbSiB layer increasing from 0 μm to 90 μm,the bending resonance frequency of the MP laminated composite gradually increases from 33.66 kHz to 39.18 kHz.This study offers a method of adjusting the strength of dc magnetic bias or the thicknesses of the FeCuNbSiB layer to tune the bending resonance frequency for ME composite,which plays a guiding role in the ME composite design for real applications.

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

Burnup measurements at the RECH-1 research reactor

International Nuclear Information System (INIS)

Henriquez, C.; Navarro, G.; Pereda, C.; Torres, H.; Pena, L.; Klein, J.; Calderon, D.; Kestelman, A.J.

2002-01-01

The Chilean Nuclear Energy Commission has decided to produce LEU fuel elements for the RECH-1 research reactor. During December 1998, the Fuel Fabrication Plant delivered the first four fuel elements, called leaders, to the RECH-1 reactor. The set was introduced into the reactor's core, following the normal routine, but performing a special follow-up on their behavior inside and outside the core. In order to measure the burn-up of the leader fuel elements, it was decided to develop a burn-up measurements system to be installed into the RECH-1 reactor pool, and to decline the use of a similar system, which operates in a hot cell. The main reason to build this facility was to have the capability to measure the burn-up of fuel elements without waiting for long decay period. This paper gives a brief description of the facility to measure the burn-up of spent fuel elements installed into the reactor pool, showing the preliminary obtained spectra and briefly discussing them. (author)

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

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.

Determination of the surface band bending in InxGa1−xN films by hard x-ray photoemission spectroscopy

Directory of Open Access Journals (Sweden)

Mickael Lozac'h, Shigenori Ueda, Shitao Liu, Hideki Yoshikawa, Sang Liwen, Xinqiang Wang, Bo Shen, Kazuaki Sakoda, Keisuke Kobayashi and Masatomo Sumiya

2013-01-01

Full Text Available Core-level and valence band spectra of InxGa1−xN films were measured using hard x-ray photoemission spectroscopy (HX-PES. Fine structure, caused by the coupling of the localized Ga 3d and In 4d with N 2s states, was experimentally observed in the films. Because of the large detection depth of HX-PES (~20 nm, the spectra contain both surface and bulk information due to the surface band bending. The InxGa1−xN films (x = 0–0.21 exhibited upward surface band bending, and the valence band maximum was shifted to lower binding energy when the mole fraction of InN was increased. On the other hand, downward surface band bending was confirmed for an InN film with low carrier density despite its n-type conduction. Although the Fermi level (EF near the surface of the InN film was detected inside the conduction band as reported previously, it can be concluded that EF in the bulk of the film must be located in the band gap below the conduction band minimum.

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

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.

HMGB1-mediated DNA bending: Distinct roles in increasing p53 binding to DNA and the transactivation of p53-responsive gene promoters.

Science.gov (United States)

Štros, Michal; Kučírek, Martin; Sani, Soodabeh Abbasi; Polanská, Eva

2018-03-01

HMGB1 is a chromatin-associated protein that has been implicated in many important biological processes such as transcription, recombination, DNA repair, and genome stability. These functions include the enhancement of binding of a number of transcription factors, including the tumor suppressor protein p53, to their specific DNA-binding sites. HMGB1 is composed of two highly conserved HMG boxes, linked to an intrinsically disordered acidic C-terminal tail. Previous reports have suggested that the ability of HMGB1 to bend DNA may explain the in vitro HMGB1-mediated increase in sequence-specific DNA binding by p53. The aim of this study was to reinvestigate the importance of HMGB1-induced DNA bending in relationship to the ability of the protein to promote the specific binding of p53 to short DNA duplexes in vitro, and to transactivate two major p53-regulated human genes: Mdm2 and p21/WAF1. Using a number of HMGB1 mutants, we report that the HMGB1-mediated increase in sequence-specific p53 binding to DNA duplexes in vitro depends very little on HMGB1-mediated DNA bending. The presence of the acidic C-terminal tail of HMGB1 and/or the oxidation of the protein can reduce the HMGB1-mediated p53 binding. Interestingly, the induction of transactivation of p53-responsive gene promoters by HMGB1 requires both the ability of the protein to bend DNA and the acidic C-terminal tail, and is promoter-specific. We propose that the efficient transactivation of p53-responsive gene promoters by HMGB1 depends on complex events, rather than solely on the promotion of p53 binding to its DNA cognate sites. Copyright © 2018 Elsevier B.V. All rights reserved.

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

Single-mode optical fiber design with wide-band ultra low bending-loss for FTTH application.

Science.gov (United States)

Watekar, Pramod R; Ju, Seongmin; Han, Won-Taek

2008-01-21

We propose a new design of a single-mode optical fiber (SMF) which exhibits ultra low bend sensitivity over a wide communication band (1.3 microm to 1.65 microm). A five-cladding fiber structure has been proposed to minimize the bending loss, estimated to be as low as 4.4x10(-10) dB/turn for the bend radius of 10 mm.

Low-bending loss and single-mode operation in few-mode optical fiber

Science.gov (United States)

Yin, Ping; Wang, Hua; Chen, Ming-Yang; Wei, Jin; Cai, Zhi-Min; Li, Lu-Ming; Yang, Ji-Hai; Zhu, Yuan-Feng

2016-10-01

The technique of eliminating the higher-order modes in a few-mode optical fiber is proposed. The fiber is designed with a group of defect modes in the cladding. The higher-order modes in the fiber can be eliminated by bending the fiber to induce strong coupling between the defect modes and the higher-order modes. Numerical simulation shows the bending losses of the LP01 mode are lower than 1.5×10-4 dB/turn for the wavelength shorter than 1.625 μm. The proposed fiber can be bent multiple turns at small bending radius which are preferable for FTTH related applications.

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.

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)

Kerr microscopy studies of the effects of bending stress on galfenola)

Science.gov (United States)

Raghunath, Ganesh; Marana, Michael; Na, Suok-Min; Flatau, Alison

2014-05-01

This work deals with using a magneto-optic Kerr effect (MOKE) microscope to optically analyze the evolution of magnetic domains in a rolled and Goss textured galfenol (Fe81Ga19 + 1.0% NbC) sample when subjected to a bending stress. The initial magnetization state of the cantilevered sample was fixed along its length by a 0.3 T permanent magnet. The magnetic state was monitored with the MOKE microscope as a tip load was applied to bend the sample. The magnetic state of galfenol depends on its magneto-elastic properties. A finite element model that incorporates an energy based formulation of magnetostriction [W. D. Armstrong, J. Magn. Magn. Mater. 263(1-2), 208-218 (2003)] was used to investigate the stresses in the sample and the corresponding change in the magnetic induction as bending occurred. A qualitative comparison with the domain pictures is presented, and the experimental micromagnetic behavior results are shown to correlate well to the macro scale bending stress and magnetization results obtained in the FEM simulations.

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)

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.

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...

The Impact of Bending Stress on the Performance of Giant Magneto-Impedance (GMI Magnetic Sensors

Directory of Open Access Journals (Sweden)

Julie Nabias

2017-03-01

Full Text Available The flexibility of amorphous Giant Magneto-Impedance (GMI micro wires makes them easy to use in several magnetic field sensing applications, such as electrical current sensing, where they need to be deformed in order to be aligned with the measured field. The present paper deals with the bending impact, as a parameter of influence of the sensor, on the GMI effect in 100 µm Co-rich amorphous wires. Changes in the values of key parameters associated with the GMI effect have been investigated under bending stress. These parameters included the GMI ratio, the intrinsic sensitivity, and the offset at a given bias field. The experimental results have shown that bending the wire resulted in a reduction of GMI ratio and sensitivity. The bending also induced a net change in the offset for the considered bending curvature and the set of used excitation parameters (1 MHz, 1 mA. Furthermore, the field of the maximum impedance, which is generally related to the anisotropy field of the wire, was increased. The reversibility and the repeatability of the bending effect were also evaluated by applying repetitive bending stresses. The observations have actually shown that the behavior of the wire under the bending stress was roughly reversible and repetitive.

Training and research on the nuclear reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.

1998-01-01

The VR-1 training reactor is a light water reactor of the pool type using enriched uranium as the fuel. The moderator is demineralized light water, which also serves as the neutron reflector, biological shielding, and coolant. Heat evolved during the fission process is removed by natural convection. The reactor is used in the education of students in the field of reactor and neutron physics, dosimetry, nuclear safety, and instrumentation and control systems for nuclear facilities. Although primarily intended for students in various branches of technology (power engineering, nuclear engineering, physical engineering), this specialized facility is also used by students of faculties educating future natural scientists and teachers. Typical tasks trained at the VR-1 reactor include: measurement of delayed neutrons; examination of the effect of various materials on the reactivity of the reactor; measurement of the neutron flux density by various procedures; measurement of reactivity by various procedures; calibration of reactor control rods by various procedures; approaching the critical state; investigation of nuclear reactor dynamics; start-up, control and operation of a nuclear reactor; and investigation of the effect of a simulated nucleate boil on reactivity. In addition to the education of university-level students, training courses are also organized for specialists in the Czech nuclear programme

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)

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.

Standard test method for guided bend test for ductility of welds

CERN Document Server

American Society for Testing and Materials. Philadelphia

2002-01-01

1.1 This test method covers a guided bend test for the determination of soundness and ductility of welds in ferrous and nonferrous products. Defects, not shown by X rays, may appear in the surface of a specimen when it is subjected to progressive localized overstressing. This guided bend test has been developed primarily for plates and is not intended to be substituted for other methods of bend testing. 1.2 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard. Note 1—For additional information see Terminology E 6, and American Welding Society Standard D 1.1. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Investigation of 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.

An alternative proposal for the disposal of potentially contaminated railroad crossties at the Savannah River Site. Revision 1

International Nuclear Information System (INIS)

Hochel, R.C.

1996-06-01

The Savannah River Site (SRS) has accumulated approximately 300,000 crossties that have been replaced over a 40-year period of maintaining the site's 62 miles of railroad track. The ties reside in a pile on an open area of 2.3 acres near the site's F-Area facilities. A small fraction of the ties are potentially contaminated with radioactivity as a result of past site practices and service. Contamination was possible from occasional leaks in transport-casks moved by rail from the site's five nuclear materials production reactors to its two reprocessing facilities. Casks typically were filled with spent fuel, targets, and water from the reactor disassembly basins, which contained small amounts of fission and neutron activation product radioactivity normal to the operation of nuclear reactors

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

Fluid-dynamic characterization of real-scale raceway reactors for microalgae production

International Nuclear Information System (INIS)

Mendoza, J.L.; Granados, M.R.; Godos, I. de; Acién, F.G.; Molina, E.; Banks, C.; Heaven, S.

2013-01-01

The fluid dynamic characterization of a 100 m length × 1 m wide channel raceway photobioreactor was carried out. The effects of water depth, liquid velocity and the presence, or absence, of sump baffles to improve the CO 2 supply transfer were considered in relation to on the power consumption, residence time and mixing in the reactor was studied. When operated at a depth of 20 cm, the power consumption was between 1.5 and 8.4 W m −3 depending on the forward velocity, with higher values occurring when the baffle was in place. Residence times and the degree of mixing at each section of the raceway (paddlewheel, bends, channels and sump) were measured experimentally. Mixing occurred mainly in the sump, paddlewheel and bends, with a maximum dispersion coefficient of 0.07 m 2 s −1 . These sections, however, only contributed a small fraction to the total volume of the raceway. Bodenstein numbers from 200 to 540 for the channel sections indicated plug-flow characteristics. Mixing times ranged from 1.4 to 6 h, with the presence of the baffle greatly increasing these times despite higher specific power consumption. A total of 15–20 circuits of the raceway were needed to achieve complete mixing without the baffle, compared to 30–40 cycles with the baffle. Vertical mixing was very poor whereas axial mixing was similar to that achieved in closed photobioreactors. The methodologies applied were shown to be useful in determining the fluid dynamics of a raceway photobioreactor. Equations useful in simulating the power consumption as a function of the design and operation parameters have been validated. -- Highlights: •Power consumption due to accessories can limit the use of raceway reactors for energy purposes. •Use of baffle to enhance mass transfer dramatically increases the power consumption in this type of photobioreactors. •High mixing time, in the order of hours, in raceway reactors limits the operation mode of these systems

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

Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

Energy Technology Data Exchange (ETDEWEB)

Otra Otra, F; Leira Rey, G

1971-07-01

In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.

Monitoring static shape memory polymers using a fiber Bragg grating as a vector-bending sensor

Science.gov (United States)

Li, Peng; Yan, Zhijun; Zhou, Kaiming; Zhang, Lin; Leng, Jinsong

2013-01-01

We propose and demonstrate a technique for monitoring the recovery deformation of the shape-memory polymers (SMP) using a surface-attached fiber Bragg grating (FBG) as a vector-bending sensor. The proposed sensing scheme could monitor the pure bending deformation for the SMP sample. When the SMP sample undergoes concave or convex bending, the resonance wavelength of the FBG will have red-shift or blue-shift according to the tensile or compressive stress gradient along the FBG. As the results show, the bending sensitivity is around 4.07 nm/cm-1. The experimental results clearly indicate that the deformation of such an SMP sample can be effectively monitored by the attached FBG not just for the bending curvature but also the bending direction.

IEA-R1 reactor - Spent fuel management

International Nuclear Information System (INIS)

Mattos, J.R.L. De

1996-01-01

Brazil currently has one Swimming Pool Research Reactor (IEA-R1) at the Instituto de Pesquisas Energeticas e Nucleares - Sao Paulo. The spent fuel produced is stored both at the Reactor Pool Storage Compartment and at the Dry Well System. The present situation and future plans for spent fuel storage are described. (author). 3 refs, 2 figs, 2 tabs

100 Area Columbia River sediment sampling

International Nuclear Information System (INIS)

Weiss, S.G.

1993-01-01

Forty-four sediment samples were collected from 28 locations in the Hanford Reach of the Columbia River to assess the presence of metals and man-made radionuclides in the near shore and shoreline settings of the Hanford Site. Three locations were sampled upriver of the Hanford Site plutonium production reactors. Twenty-two locations were sampled near the reactors. Three locations were sampled downstream of the reactors near the Hanford Townsite. Sediment was collected from depths of 0 to 6 in. and between 12 to 24 in. below the surface. Samples containing concentrations of metals exceeding the 95 % upper threshold limit values (DOE-RL 1993b) are considered contaminated. Contamination by arsenic, chromium, copper, lead, and zinc was found. Man-made radionuclides occur in all samples except four collected opposite the Hanford Townsite. Man-made radionuclide concentrations were generally less than 1 pCi/g

100 Area Columbia River sediment sampling

Energy Technology Data Exchange (ETDEWEB)

Weiss, S.G. [Westinghouse Hanford Co., Richland, WA (United States)

1993-09-08

Forty-four sediment samples were collected from 28 locations in the Hanford Reach of the Columbia River to assess the presence of metals and man-made radionuclides in the near shore and shoreline settings of the Hanford Site. Three locations were sampled upriver of the Hanford Site plutonium production reactors. Twenty-two locations were sampled near the reactors. Three locations were sampled downstream of the reactors near the Hanford Townsite. Sediment was collected from depths of 0 to 6 in. and between 12 to 24 in. below the surface. Samples containing concentrations of metals exceeding the 95 % upper threshold limit values (DOE-RL 1993b) are considered contaminated. Contamination by arsenic, chromium, copper, lead, and zinc was found. Man-made radionuclides occur in all samples except four collected opposite the Hanford Townsite. Man-made radionuclide concentrations were generally less than 1 pCi/g.

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....

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)

TR-EDB: Test Reactor Embrittlement Data Base, Version 1

Energy Technology Data Exchange (ETDEWEB)

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

1994-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

TR-EDB: Test Reactor Embrittlement Data Base, Version 1

International Nuclear Information System (INIS)

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

1994-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

Modernization and Refurbishment of the RECH-1 Reactor

Energy Technology Data Exchange (ETDEWEB)

Daie, J. [Nuclear Application Department, Chilean Nuclear Energy Commission (CCHEN), Santiago (Chile)

2014-08-15

The Chilean Nuclear Energy Commission (Comisión Chilena de Energía Nuclear, or CCHEN) has operated the RECH-1 research reactor since 1974. This reactor is located at La Reina Nuclear Centre in Santiago, Chile. It is a pool type reactor using LEU MTR fuel assemblies, light water as moderator and coolant, and beryllium as reflector. The reactor has been operated at the nominal power of 5 MW in a continuous shift of 20 hours per week, 48 weeks per year. The main utilizations of the RECH-1 reactor are radioisotope production and neutron activation analysis. Among the most relevant refurbishment and modernization campaigns undertaken at the reactor are: full core conversion to the use of LEU fuel, replacement of the cooling tower, improvement of the containment building by changing the doors and gates and by a better sealant for the penetrations, introduction of an additional source of water by connecting the raw water supply system to the emergency cooling system, improvement of the emergency ventilation system, introduction of a fire detection and alarm system for detection and mitigation to protect the I&C racks, introduction of a radioactive liquid release for those generated at the reactor, introduction of a delay tank degasification system and renewal of the environmental monitoring system. At present we are assessing the possibility of replacing the old analog electronics of control for new digital systems. Detailed descriptions of these diverse activities are presented in the paper. (author)

Involvement of Sodium Nitroprusside (SNP in the Mechanism That Delays Stem Bending of Different Gerbera Cultivars

Directory of Open Access Journals (Sweden)

Aung H. Naing

2017-11-01

Full Text Available Longevity of cut flowers of many gerbera cultivars (Gerbera jamesonii is typically short because of stem bending; hence, stem bending that occurs during the early vase life period is a major problem in gerbera. Here, we investigated the effects of sodium nitroprusside (SNP on the delay of stem bending in the gerbera cultivars, Alliance, Rosalin, and Bintang, by examining relative fresh weight, bacterial density in the vase solution, transcriptional analysis of a lignin biosynthesis gene, antioxidant activity, and xylem blockage. All three gerbera cultivars responded to SNP by delaying stem bending, compared to the controls; however, the responses were dose- and cultivar-dependent. Among the treatments, SNP at 20 mg L-1 was the best to delay stem bending in Alliance, while dosages of 10 and 5 mg L-1 were the best for Rosalin and Bintang, respectively. However, stem bending in Alliance and Rosalin was faster than in Bintang, indicating a discrepancy influenced by genotype. According to our analysis of the role of SNP in the delay of stem bending, the results revealed that SNP treatment inhibited bacterial growth and xylem blockage, enhanced expression levels of a lignin biosynthesis gene, and maintained antioxidant activities. Therefore, it is suggested that the cause of stem bending is associated with the above-mentioned parameters and SNP is involved in the mechanism that delays stem bending in the different gerbera cultivars.

Extensive utilisation of VR-1 reactor for nuclear education and training

International Nuclear Information System (INIS)

Rataj, J.

2010-01-01

The paper presents utilisation of the VR-1 reactor for nuclear education and training at national and international level. VR-1 reactor has been operating by the Czech Technical University since December 1990. The reactor is a pool-type light water reactor based on enriched uranium (19.7% 235 U) with maximum thermal power 1kW and for short time period up to 5kW. The moderator of neutrons is light water, which is also used as a reflector, a biological shielding and a coolant. Heat is removed from the core by natural convection. The pool disposition of the reactor facilitates access to the core, setting and removing of various experimental samples and detectors, easy and safe handling of fuel assemblies. The reactor core can contain from 17 to 21 fuel assemblies IRT-4M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The reactor is equipped with several experimental devices; e.g. horizontal, radial and tangential channels used to take out a neutron beam, reactivity oscillator for dynamics study and bubble boiling simulator. The reactor has been used very efficiently especially for education and training of university students and NPP's specialists for more than 18 years. The VR-1 reactor is utilised within various national and international activities such as Czech Nuclear Education Network (CENEN), European Nuclear Education Network and also Eastern European Research Reactor Initiative (EERRI). The reactor is well equipped for education and training not only by the experimental facility itself but also by incessant development of training methods and improvement of education experiments. The education experiments can be combined into training courses attended by students according to their study specialization and knowledge level. The training programme is aimed to the reactor and neutron physics, dosimetry, nuclear safety, and control of nuclear installations. Every year, approximately 250 university students undergo

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

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

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)

Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

KAUST Repository

Zhao, Zhenlong

2013-01-17

Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

Use of the VR-1 ''Vrabec'' training reactor

International Nuclear Information System (INIS)

Matejka, K.; Kolros, A.; Krops, S.; Polach, S.; Sklenka, L.

1994-01-01

An overview is presented of the extent and ways of using the VR-1 training reactor, which is operated by the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague. A list and the characteristics of 16 problems developed for teaching purposes is given, and the 14 faculties and 2 research institutes participating in the teaching activities are listed. The reactor is used in the education and training of nuclear scientists and engineers. The instrumentation, experimental, handling and operating tools, as well as documentation and texts relating to the reactor are described. The following examples of the teaching activities are included: a guided visit to the operating reactor site, reactor dynamics study and delayed neutron measurement, training course, and the basic criticality experiment. Nuclear safety aspects (hypothetical accidents, quality control and system qualification demonstration, safety culture) are stressed during the education. The reactor department is involved in international cooperation projects. (J.B.). 3 refs

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

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

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.

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

Electrical system regulations of the IEA-R1 reactor

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2013-01-01

The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

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)

Numerical investigation into strong axis bending-shear interaction in rolled I-shaped steel sections

NARCIS (Netherlands)

Dekker, R.W.A.; Snijder, H.H.; Maljaars, J.; Dubina, Dan; Ungureanu, Viorel

2016-01-01

Clause 6.2.8 of EN 1993-1-1 covers the design rules on bending-shear resistance, taking presence of shear into account by a reduced yield stress for the shear area. Numerical research on bending-shear interaction by means of the Abaqus Finite Element modelling software is presented. The numerical

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)

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...

Modernization of control instrumentation and security of reactor IAN - R1

International Nuclear Information System (INIS)

Gonzalez, J. M.

1993-01-01

The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels

Operation and maintenance of 1MW PUSPATI TRIGA reactor

International Nuclear Information System (INIS)

Adnan Bokhari; Mohammad Suhaimi Kassim

2006-01-01

The Malaysian Research Reactor, Reactor TRIGA PUSPATI (RTP) has been successfully operated for 22 years for various experiments. Since its commissioning in June 1982 until December 2004, the 1MW pool-type reactor has accumulated more than 21143 hours of operation, corresponding to cumulative thermal energy release of about 14083 MW-hours. The reactor is currently in operation and normally operates on demand, which is normally up to 6 hours a day. Presently the reactor core is made up of standard TRIAGA fuel element consists of 8.5 wt%, 12 wt% and 20 wt% types; 20%-enriched and stainless steel clad. Several measures such as routine preventive maintenance and improving the reactor support systems have been taken toward achieving this long successful operation. Besides normal routine utilization like other TRIGA reactors, new strategies are implemented for effective increase in utilization. (author)

33 CFR 223.1 - Mississippi River Water Control Management Board.

Science.gov (United States)

2010-07-01

..., responsibilities and authority of the Mississippi River Water Control Management Board. (b) Applicability. This... control management within the Mississippi River Basin. (c) Objectives. The objectives of the Board are: (1...) Composition. The Mississippi River Water Control Management Board is a continuing board consisting of the...

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

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

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

Liquid-metal-cooled, curved-crystal monochromator for Advanced Photon Source bending-magnet beamline 1-BM

International Nuclear Information System (INIS)

Brauer, S.; Rodricks, B.; Assoufid, L.; Beno, M.A.; Knapp, G.S.

1996-06-01

The authors describe a horizontally focusing curved-crystal monochromator that invokes a 4-point bending scheme and a liquid-metal cooling bath. The device has been designed for dispersive diffraction and spectroscopy in the 5--20 keV range, with a predicted focal spot size of ≤ 100 microm. To minimize thermal distortions and thermal equilibration time, the 355 x 32 x 0.8 mm crystal will be nearly half submerged in a bath of Ga-In-Sn-Zn alloy. The liquid metal thermally couples the crystal to the water-cooled Cu frame, while permitting the required crystal bending. Calculated thermal profiles and anticipated focusing properties are discussed

A study on the impulse wave discharged from the exit of a right-angle pipe bend

International Nuclear Information System (INIS)

Lee, D. H.; Hur, S. C.; Kweon, Y. H.; Kim, H. D.

2001-01-01

The current study addresses experimental and computational work of impulse wave discharged from the exit of two kinds of right-angle pipe bends, which are attached to the open end of a simple shock tube. The weak normal shock wave with its magnitude of Mach number from 1.02 to 1.20 is employed to obtain the impulse wave propagating outside the exit of the pipe bends. A Schlieren optical system visualizes the impulse wave discharged from the exit of the pipe bends at an instant. The experimental data of the magnitude of the impulse wave and its propagating directivity are analyzed to characterize the impulse waves discharged from the exit of the pipe bends and compared with those discharged from a straight pipe. Computational results well predict the experimented dynamic behaviors of the impulse wave. The results obtained show that a right-angle miter bend considerably reduces the magnitude of the impulse wave and its directivity toward to the pipe axis, compared with the straight pipe and right-angle smooth bend. It is believed that the right-angle miter bend pipe can play a role of a passive control against the impulse wave

Perspectives on reactor safety. Revision 1

International Nuclear Information System (INIS)

Haskin, F.E.; Hodge, S.A.

1997-11-01

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

Perspectives on reactor safety. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.

1997-11-01

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

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

Numerical investigation into strong axis bending shear interaction in rolled I-shaped steel sections

NARCIS (Netherlands)

Dekker, R.W.A.; Snijder, B.H.; Maljaars, J.

2016-01-01

Clause 6.2.8 of EN 1993-1-1 covers the design rules on bending-shear resistance, taking presence of shear into account by a reduced yield stress for the shear area. Numerical research on bending-shear interaction by means of the Abaqus Finite Element modelling soft-ware is presented. The numerical

Colloid-Colloid Hydrodynamic Interaction Around a Bend in a Quasi-One-Dimensional Channel

Science.gov (United States)

Liepold, Christopher; Zarcone, Ryan; Heumann, Tibor; Lin, Binhua; Rice, Stuart

We report a study of the correlation between a pair of particles in a colloid suspension in a bent quasi-one-dimensional (q1d) channel as a function of bend angle. As the bend angle becomes more acute, we observe an increasing depletion of particles in the vicinity of the bend and an increase in the nearest-neighbor separation in the pair correlation function for particles on opposite sides of the bend. Further, we observe that the peak value of D12, the coupling term in the pair diffusion tensor that characterizes the effect of the motion of particle 1 on particle 2, coincides with the first peak in the pair correlation function, and that the pair separation dependence of D12 mimics that of the pair correlation function. We show that the observed behavior is a consequence of the geometric constraints imposed by the single-file requirement that the particle centers lie on the centerline of the channel and the requirement that the hydrodynamic flow must follow the channel around the bend. We find that the correlation between a pair of particles in a colloidal suspension in a bent q1D channel has the same functional dependence on the pair correlation function as in a straight q1D channel when measured in a coordinate system that follows the centerline of the bent channel. NSF MRSEC (DMR-1420709), Dreyfus Foundation (SI-14-014).

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

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 .

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).

Annual report on JEN-1 and JEN-2 Reactors

International Nuclear Information System (INIS)

Montes Ponce de Leon, J.

1974-01-01

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)

Advances in Reactor Physics, Mathematics and Computation. Volume 1

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume one, are divided into 6 sessions bearing on: - session 1: Advances in computational methods including utilization of parallel processing and vectorization (7 conferences) - session 2: Fast, epithermal, reactor physics, calculation, versus measurements (9 conferences) - session 3: New fast and thermal reactor designs (9 conferences) - session 4: Thermal radiation and charged particles transport (7 conferences) - session 5: Super computers (7 conferences) - session 6: Thermal reactor design, validation and operating experience (8 conferences).

Stationary low power reactor No. 1 (SL-1) accident site decontamination ampersand dismantlement project

International Nuclear Information System (INIS)

Perry, E.F.

1995-01-01

The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the building contained the reactor pressure vessel surrounded by gravel shielding. Above the pressure vessel, in the center portion of the building, was a turbine generator and plant support equipment. The upper section of the building contained an air cooled condenser and its circulation fan. The major support facilities included a 2,500 ft 2 two story, cinder block administrative building; two 4,000 ft 2 single story, steel frame office buildings; a 850 ft 2 steel framed, metal sided PL condenser building, and a 550 ft 2 steel framed decontamination and laydown building

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

Reactor utilization, Part 1

International Nuclear Information System (INIS)

Martinc, R.; Stanic, A.

1981-01-01

The reactor operating plan for 1981 was subject to the needs of testing operation with the 80% enriched fuel and was fulfilled on the whole. This annex includes data about reactor operation, review of shorter interruptions due to demands of the experiments, data about safety shutdowns caused by power cuts. Period of operation at low power levels was used mostly for activation analyses, and the operation at higher power levels were used for testing and regular isotope production. Detailed data about samples activation are included as well as utilization of the reactor as neutron source and the operating plan for 1982 [sr

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

Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

International Nuclear Information System (INIS)

Salahuddin, A.; Israr, M.; Hussain, M.

2002-01-01

Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)

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...

Second-order infinitesimal bendings of surfaces of revolution with flattening at the poles

International Nuclear Information System (INIS)

Sabitov, I Kh

2014-01-01

We study infinitesimal bendings of surfaces of revolution with flattening at the poles. We begin by considering the minimal possible smoothness class C 1 both for surfaces and for deformation fields. Conditions are formulated for a given harmonic of a first-order infinitesimal bending to be extendable into a second order infinitesimal bending. We finish by stating a criterion for nonrigidity of second order for closed surfaces of revolution in the analytic class. We also give the first concrete example of such a nonrigid surface. Bibliography: 15 entries

Second-order infinitesimal bendings of surfaces of revolution with flattening at the poles

Energy Technology Data Exchange (ETDEWEB)

Sabitov, I Kh [M. V. Lomonosov Moscow State University, Faculty of Mechanics and Mathematics, Moscow (Russian Federation)

2014-12-31

We study infinitesimal bendings of surfaces of revolution with flattening at the poles. We begin by considering the minimal possible smoothness class C{sup 1} both for surfaces and for deformation fields. Conditions are formulated for a given harmonic of a first-order infinitesimal bending to be extendable into a second order infinitesimal bending. We finish by stating a criterion for nonrigidity of second order for closed surfaces of revolution in the analytic class. We also give the first concrete example of such a nonrigid surface. Bibliography: 15 entries.

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

RA research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1985

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1985-01-01

According to the plan, RA reactor was to be in operation in mid September 1985. But, since the building of the emergency cooling system, nor the reconstruction of the existing special ventilation system were not finished until the end of August reactor was not operated during 1985. During the previous four years reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care, which was cancelled in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981-1984. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks have started: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. IAEA has approved the amount of 1,300,000 US dollars for the renewal of the instrumentation [sr

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

New instrumentation for the IPR-R1 reactor of CDTN

International Nuclear Information System (INIS)

Carvalho, P.V.R. de.

1992-01-01

The Nuclear Engineering Institute reactor instrumentation area has developed systems and equipment for reactor operation and safety. In such way, the new I and C for IEN Argonauta reactor and the nuclear instrumentation for IPEN critical facility were built. This paper describes our real work, the new I and C systems for IPR-R1, a Triga type reactor, located at CDTN (Belo Horizonte - MG). (author)

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.

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

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.

METHYL JASMONATE AND STEM BENDING HARDENING AND INITIAL GROWTH OF Cordia trichotoma SEEDLINGS

Directory of Open Access Journals (Sweden)

Danielle Acco Cadorin

2015-12-01

Full Text Available The submission of seedlings to mechanical stimuli and plant growth regulator promote their hardening and can be included in the routine of nurseries, favoring the survival and initial growth in the field. The study aimed to evaluate the effects of applying methyl jasmonate and stem bending in hardening and initial growth of Cordia trichotoma seedlings. Seedlings were subjected to 20 stem bending daily for 4 weeks; 20 stem bending daily for 8 weeks; 50 µmol.L-1 of methyl jasmonate applied weekly for 4 weeks; 50 µmol.L-1 of methyl jasmonate applied weekly for 8 weeks and the control treatment. The design was a completely randomized, with five repetitions of the fourteen seedlings. Seedlings submitted to hardening treatments showed less increment in height, greater increment in stem diameter and less value for strength index. Seedlings of control treatment had greater loss of root tissue electrolytes and less potential for root regeneration. In the field, 180 days after planting, seedlings submitted to eight weeks of stem bending and eight methyl jasmonate applications showed greater increment in height and stem diameter. The results indicate that both stem bending such as methyl jasmonate application for eight weeks are effective in promoting hardening and improve the starting performance in field of Cordia trichotoma seedlings.

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)

All-fiber intensity bend sensor based on photonic crystal fiber with asymmetric air-hole structure

Science.gov (United States)

Budnicki, Dawid; Szostkiewicz, Lukasz; Szymanski, Michal O.; Ostrowski, Lukasz; Holdynski, Zbigniew; Lipinski, Stanislaw; Murawski, Michal; Wojcik, Grzegorz; Makara, Mariusz; Poturaj, Krzysztof; Mergo, Pawel; Napierala, Marek; Nasilowski, Tomasz

2017-10-01

Monitoring the geometry of an moving element is a crucial task for example in robotics. The robots equipped with fiber bend sensor integrated in their arms can be a promising solution for medicine, physiotherapy and also for application in computer games. We report an all-fiber intensity bend sensor, which is based on microstructured multicore optical fiber. It allows to perform a measurement of the bending radius as well as the bending orientation. The reported solution has a special airhole structure which makes the sensor only bend-sensitive. Our solution is an intensity based sensor, which measures power transmitted along the fiber, influenced by bend. The sensor is based on a multicore fiber with the special air-hole structure that allows detection of bending orientation in range of 360°. Each core in the multicore fiber is sensitive to bend in specified direction. The principle behind sensor operation is to differentiate the confinement loss of fundamental mode propagating in each core. Thanks to received power differences one can distinguish not only bend direction but also its amplitude. Multicore fiber is designed to utilize most common light sources that operate at 1.55 μm thus ensuring high stability of operation. The sensitivity of the proposed solution is equal 29,4 dB/cm and the accuracy of bend direction for the fiber end point is up to 5 degrees for 15 cm fiber length. Such sensitivity allows to perform end point detection with millimeter precision.

Performance of composite I-beams under axial compression and bending load modes

International Nuclear Information System (INIS)

Khalid, Y.A.; Ali, F.A.; Sahari, B.B.; Saad, E.M.A.

2005-01-01

An experimental and finite-element analyses for glass/epoxy composite I-beams have been carried out. Four, six, eight and 10 layers of woven fabric glass/epoxy composite I-beams were fabricated by a hand lay-up (molding) process. Quasi-static axial crushing and bending loading modes were used for this investigation. The load-displacement response was obtained and the energy absorption values were calculated for all the composite I-beams. Three tests were done for each composite I-beams type and each loading case for the results conformation. The second part of this study includes the elastic behavior of composite I-beams of the same dimensions and materials using finite-element analysis. The woven fabric glass/epoxy composite I-beams mechanical properties have been obtained from tensile tests. Results from this investigation show that the load required and the specific energy absorption for composite I-beams under axial compression load were higher than those for three and four point bending. On the other hand, the loads required for composite I-beams under four point bending were higher than those for three point bending, while the specific energy absorption for composite I-beams under three point bending were higher than those for four point bending. The first crushing loads difference between the experimental and finite-element results fell in the 3.6-10.92% range for axial compression tests, while fell in the 1.44-12.99% and 4.94-22.0% range for three and four point bending, respectively

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

Progress on traveling-wave reactor design

International Nuclear Information System (INIS)

Gilleland, John

2009-01-01

TerraPower LLC is leading a collaborative effort to develop physics and engineering designs for several kinds of sodium-cooled traveling-wave reactors. This collaboration includes nuclear engineering groups at TerraPower, M.I.T., U.N.L.V., Argonne National Laboratory, and the Columbia River Basin Consulting Group, as well as individual consultants from Lawrence Livermore National Laboratory, U.C. Berkeley, and several other institutions. The goal of this initiative is to develop innovative technologies that will enable cost-effective breed-and-burn reactors, which produce electricity from fuel composed almost wholly of depleted uranium. We will present conceptual designs ranging in reactor vessel size from five meters to 13 meters and in output from about 100 MWe to more than 1,000 MWe. Our Monte Carlo simulations for these reactors predict refueling intervals ranging from 40 to 125 years. Scaling designs from small to large sizes requires a shift in basic design approach; lessons learned from this effort will be discussed. We will also share our evolving understanding of the ways in which the core design can be simplified by improvements to certain limiting technologies. (author)

Influence of flock coating on bending rigidity of woven fabrics

Science.gov (United States)

Ozdemir, O.; Kesimci, M. O.

2017-10-01

This work presents the preliminary results of our efforts that focused on the effect of the flock coating on the bending rigidity of woven fabrics. For this objective, a laboratory scale flocking unit is designed and flocked samples of controlled flock density are produced. Bending rigidity of the samples with different flock densities are measured on both flocked and unflocked sides. It is shown that the bending rigidity depends on both flock density and whether the side to be measured is flocked or not. Adhesive layer thickness on the bending rigidity is shown to be dramatic. And at higher basis weights, flock density gets less effective on bending rigidity.

Response of the St. Joseph River to lake level changes during the last 12,000 years in the Lake Michigan basin

Science.gov (United States)

Kincare, K.A.

2007-01-01

The water level of the Lake Michigan basin is currently 177 m above sea level. Around 9,800 14C years B.P., the lake level in the Lake Michigan basin had dropped to its lowest level in prehistory, about 70 m above sea level. This low level (Lake Chippewa) had profound effects on the rivers flowing directly into the basin. Recent studies of the St. Joseph River indicate that the extreme low lake level rejuvenated the river, causing massive incision of up to 43 m in a valley no more than 1.6 km wide. The incision is seen 25 km upstream of the present shoreline. As lake level rose from the Chippewa low, the St. Joseph River lost competence and its estuary migrated back upstream. Floodplain and channel sediments partially refilled the recently excavated valley leaving a distinctly non-classical morphology of steep sides with a broad, flat bottom. The valley walls of the lower St. Joseph River are 12-18 m tall and borings reveal up to 30 m of infill sediment below the modern floodplain. About 3 ?? 108 m3 of sediment was removed from the St. Joseph River valley during the Chippewa phase lowstand, a massive volume, some of which likely resides in a lowstand delta approximately 30 km off-shore in Lake Michigan. The active floodplain below Niles, Michigan, is inset into an upper terrace and delta graded to the Calumet level (189 m) of Lake Chicago. In the lower portion of the terrace stratigraphy a 1.5-2.0 m thick section of clast-supported gravel marks the entry of the main St. Joseph River drainage above South Bend, Indiana, into the Lake Michigan basin. This gravel layer represents the consolidation of drainage that probably occurred during final melting out of ice-marginal kettle chains allowing stream piracy to proceed between Niles and South Bend. It is unlikely that the St. Joseph River is palimpsest upon a bedrock valley. The landform it cuts across is a glaciofluvial-deltaic feature rather than a classic unsorted moraine that would drape over pre-glacial topography

Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1

Energy Technology Data Exchange (ETDEWEB)

Tippayakul, Chanatip [Thailand Institute of Nuclear Technology, Bangkok (Thailand)

2013-07-01

Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was

Flexible robotic entry device for a nuclear materials production reactor

International Nuclear Information System (INIS)

Heckendorn, F.M. II.

1988-01-01

The Savannah River Laboratory has developed and is implementing a flexible robotic entry device (FRED) for the nuclear materials production reactors now operating at the Savannah River Plant (SRP). FRED is designed for rapid deployment into confinement areas of operating reactors to assess unknown conditions. A unique smart tether method has been incorporated into FRED for simultaneous bidirectional transmission of multiple video/audio/control/power signals over a single coaxial cable. This system makes it possible to use FRED under all operating and standby conditions, including those where radio/microwave transmissions are not possible or permitted, and increases the quantity of data available

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...

Hydrodynamic processes in sharp meander bends and their morphological implications

NARCIS (Netherlands)

Blanckaert, K.

2011-01-01

The migration rate of sharp meander bends exhibits large variance and indicates that some sharply curved bends tend to stabilize. These observations remain unexplained. This paper examines three hydrodynamic processes in sharp bends with fixed banks and discusses their morphological implications:

Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

International Nuclear Information System (INIS)

Veloso, Marcelo Antonio; Fortini, Maria Auxiliadora

2002-01-01

The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

Environmental Information Document: L-reactor reactivation

International Nuclear Information System (INIS)

Mackey, H.E. Jr.

1982-04-01

Purpose of this Environmental Information Document is to provide background for assessing environmental impacts associated with the renovation, restartup, and operation of L Reactor at the Savannah River Plant (SRP). SRP is a major US Department of Energy installation for the production of nuclear materials for national defense. The purpose of the restart of L Reactor is to increase the production of nuclear weapons materials, such as plutonium and tritium, to meet projected needs in the nuclear weapons program

Environmental Information Document: L-reactor reactivation

Energy Technology Data Exchange (ETDEWEB)

Mackey, H.E. Jr. (comp.)

1982-04-01

Purpose of this Environmental Information Document is to provide background for assessing environmental impacts associated with the renovation, restartup, and operation of L Reactor at the Savannah River Plant (SRP). SRP is a major US Department of Energy installation for the production of nuclear materials for national defense. The purpose of the restart of L Reactor is to increase the production of nuclear weapons materials, such as plutonium and tritium, to meet projected needs in the nuclear weapons program.

Statistical Analysis of Bending Rigidity Coefficient Determined Using Fluorescence-Based Flicker-Noise Spectroscopy.

Science.gov (United States)

Doskocz, Joanna; Drabik, Dominik; Chodaczek, Grzegorz; Przybyło, Magdalena; Langner, Marek

2018-06-01

Bending rigidity coefficient describes propensity of a lipid bilayer to deform. In order to measure the parameter experimentally using flickering noise spectroscopy, the microscopic imaging is required, which necessitates the application of giant unilamellar vesicles (GUV) lipid bilayer model. The major difficulty associated with the application of the model is the statistical character of GUV population with respect to their size and the homogeneity of lipid bilayer composition, if a mixture of lipids is used. In the paper, the bending rigidity coefficient was measured using the fluorescence-enhanced flicker-noise spectroscopy. In the paper, the bending rigidity coefficient was determined for large populations of 1-palmitoyl-2-oleoyl-sn-glycero-3-phosphocholine and 1,2-dioleoyl-sn-glycero-3-phosphocholine vesicles. The quantity of obtained experimental data allows to perform statistical analysis aiming at the identification of the distribution, which is the most appropriate for the calculation of the value of the membrane bending rigidity coefficient. It has been demonstrated that the bending rigidity coefficient is characterized by an asymmetrical distribution, which is well approximated with the gamma distribution. Since there are no biophysical reasons for that we propose to use the difference between normal and gamma fits as a measure of the homogeneity of vesicle population. In addition, the effect of a fluorescent label and types of instrumental setups on determined values has been tested. Obtained results show that the value of the bending rigidity coefficient does not depend on the type of a fluorescent label nor on the type of microscope used.

Response Matrix Method Development Program at Savannah River Laboratory

International Nuclear Information System (INIS)

Sicilian, J.M.

1976-01-01

The Response Matrix Method Development Program at Savannah River Laboratory (SRL) has concentrated on the development of an effective system of computer codes for the analysis of Savannah River Plant (SRP) reactors. The most significant contribution of this program to date has been the verification of the accuracy of diffusion theory codes as used for routine analysis of SRP reactor operation. This paper documents the two steps carried out in achieving this verification: confirmation of the accuracy of the response matrix technique through comparison with experiment and Monte Carlo calculations; and establishment of agreement between diffusion theory and response matrix codes in situations which realistically approximate actual operating conditions

Summary of failed reactor coolant pump rotating assembly experience at Crystal River Unit 3

International Nuclear Information System (INIS)

Hayner, G.O.; Clary, M.D.

1992-01-01

Four reactor coolant pump (RCP) rotating assemblies (shafts) have failed or have severely cracked during operation at the Crystal River Unit 3 (CR-3) Nuclear Power Plant. The two failed shafts removed from RCP-1A have been extensively examined. All of the RCP shafts (except the D shaft) were fabricated from UNS S66286 superalloy (Alloy A-286). The D shaft was fabricated from UNS S20910 (Alloy XM-19/Nitronic 50). Torsional strain gauge analysis was performed on the RCP-1A shaft during the 1990 refueling outage. This type of analysis has not been performed previously on an operating RCP. Several results were found including: (1) the primary components of alternating torsional stress during normal RCP operation are impeller vane pass and a sub-2X torsional resonance with maximum components of ∼±0.8 ksi; (2) a typical vane pass cycle is initiated by an abrupt unloading of the shaft followed by a reload past equilibrium and a damped return to equilibrium; (3) a higher (compared to normal four pump operation) alternating torsional stress range resulted from solo operation of RCP-1A at low temperature and pressure (normal startup conditions); (4) the 2/0 combination produced the highest mean torsional stresses and the lowest alternating stresses and (5) a startup of a secured RCP with three operating pumps results in significantly higher alternating stress than a cold startup. The root cause RCP failure mechanism appears to involve RCP startup sequence at CR-3, peculiarities that necessitate this sequence and complex shaft stresses just above or under the journal bearing. The 1986 impeller bolt failure is not considered to be a root cause effect. It was also determined that fatigue cracking has always been responsible for both shaft initiation and propagation mechanisms and cracking can occur independent of shaft material

New human machine interface for VR-1 training reactor

International Nuclear Information System (INIS)

Kropik, M.; Matejka, K.; Sklenka, L.; Chab, V.

2002-01-01

The contribution describes a new human machine interface that was installed at the VR-1 training reactor. The human machine interface update was completed in the summer 2001. The human machine interface enables to operate the training reactor. The interface was designed with respect to functional, ergonomic and aesthetic requirements. The interface is based on a personal computer equipped with two displays. One display enables alphanumeric communication between a reactor operator and the control and safety system of the nuclear reactor. Messages appear from the control system, the operator can write commands and send them there. The second display is a graphical one. It is possible to represent there the status of the reactor, principle parameters (as power, period), control rods' positions, the course of the reactor power. Furthermore, it is possible to set parameters, to show the active core configuration, to perform reactivity calculations, etc. The software for the new human machine interface was produced in the InTouch developing environment of the WonderWare Company. It is possible to switch the language of the interface between Czech and English because of many foreign students and visitors at the reactor. The former operator's desk was completely removed and superseded with a new one. Besides of the computer and the two displays, there are control buttons, indicators and individual numerical displays of instrumentation there. Utilised components guarantee high quality of the new equipment. Microcomputer based communication units with proper software were developed to connect the contemporary control and safety system with the personal computer of the human machine interface and the individual displays. New human machine interface at the VR-1 training reactor improves the safety and comfort of the reactor utilisation, facilitates experiments and training, and provides better support of foreign visitors.(author)

Safety issues at the defense production reactors

International Nuclear Information System (INIS)

1987-01-01

The United States produces plutonium and tritium for use in nuclear weapons at the defense production reactors - the N Reactor in Washington and the Savannah River reactors in South Carolina. This report reaches general conclusions about the management of those reactors and highlights a number of safety and technical issues that should be resolved. The report provides an assessment of the safety management, safety review, and safety methodology employed by the Department of Energy and the private contractors who operate the reactors for the federal government. This report examines the safety objective established by the Department of Energy for the production reactors and the process the Department of its contractors use to implement the objective; focuses on a variety of uncertainties concerning the production reactors, particularly those related to potential vulnerabilities to severe accidents; and identifies ways in which the DOE approach to management of the safety of the production reactors can be improved

Bending spring rate investigation of nanopipette for cell injection

Science.gov (United States)

Shen, Yajing; Zhang, Zhenhai; Fukuda, Toshio

2015-04-01

Bending of nanopipette tips during cell penetration is a major cause of cell injection failure. However, the flexural rigidity of nanopipettes is little known due to their irregular structure. In this paper, we report a quantitative method to estimate the flexural rigidity of a nanopipette by investigating its bending spring rate. First nanopipettes with a tip size of 300 nm are fabricated from various glass tubes by laser pulling followed by focused ion beam (FIB) milling. Then the bending spring rate of the nanopipettes is investigated inside a scanning electron microscope (SEM). Finally, a yeast cell penetration test is performed on these nanopipettes, which have different bending spring rates. The results show that nanopipettes with a higher bending spring rate have better cell penetration capability, which confirms that the bending spring rate may well reflect the flexural rigidity of a nanopipette. This method provides a quantitative parameter for characterizing the mechanical property of a nanopipette that can be potentially taken as a standard specification in the future. This general method can also be used to estimate other one-dimensional structures for cell injection, which will greatly benefit basic cell biology research and clinical applications.

Bending spring rate investigation of nanopipette for cell injection

International Nuclear Information System (INIS)

Shen, Yajing; Zhang, Zhenhai; Fukuda, Toshio

2015-01-01

Bending of nanopipette tips during cell penetration is a major cause of cell injection failure. However, the flexural rigidity of nanopipettes is little known due to their irregular structure. In this paper, we report a quantitative method to estimate the flexural rigidity of a nanopipette by investigating its bending spring rate. First nanopipettes with a tip size of 300 nm are fabricated from various glass tubes by laser pulling followed by focused ion beam (FIB) milling. Then the bending spring rate of the nanopipettes is investigated inside a scanning electron microscope (SEM). Finally, a yeast cell penetration test is performed on these nanopipettes, which have different bending spring rates. The results show that nanopipettes with a higher bending spring rate have better cell penetration capability, which confirms that the bending spring rate may well reflect the flexural rigidity of a nanopipette. This method provides a quantitative parameter for characterizing the mechanical property of a nanopipette that can be potentially taken as a standard specification in the future. This general method can also be used to estimate other one-dimensional structures for cell injection, which will greatly benefit basic cell biology research and clinical applications. (paper)

Evaluation of spinal instrumentation rod bending characteristics for in-situ contouring.

Science.gov (United States)

Noshchenko, Andriy; Xianfeng, Yao; Armour, Grant Alan; Baldini, Todd; Patel, Vikas V; Ayers, Reed; Burger, Evalina

2011-07-01

Bending characteristics were studied in rods used for spinal instrumentation at in-situ contouring conditions. Five groups of five 6 mm diameter rods made from: cobalt alloy (VITALLIUM), titanium-aluminum-vanadium alloy (SDI™), β-titanium alloy (TNTZ), cold worked stainless steel (STIFF), and annealed stainless steel (MALLEABLE) were studied. The bending procedure was similar to that typically applied for in-situ contouring in the operating room and included two bending cycles: first--bending to 21-24° under load with further release of loading for 10 min, and second--bending to 34-37° at the previously bent site and release of load for 10 min. Applied load, bending stiffness, and springback effect were studied. Statistical evaluation included ANOVA, correlation and regression analysis. TNTZ and SDI™ rods showed the highest (p under load (p < 0.001). To reach the necessary bend angle after unloading, over bending should be 37-40% of the required angle in TNTZ and SDI™ rods, 27-30% in VITALLIUM and STIFF rods, and around 20% in MALLEABLE rods. Copyright © 2011 Wiley Periodicals, Inc.

Reactor Engineering Department annual report (April 1, 1987 - March 31, 1988)

International Nuclear Information System (INIS)

1988-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1987 (April 1, 1987 - March 31, 1988). The major activities in the Department concerns the programs of the high temperature gas-cooled reactor, the high conversion light water reactor, the advanced fission reactor system and the fusion reactor at JAERI and the fast breeder reactor at PNC. The report contains the latest progress in nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control/diagnosis and robotics, as well as the new topics from this fiscal year on advanced reactors system design studies and technique developments related the facilities in the Department. Also described are the activities of the Research Committee on Reactor Physics. (author)