Develoment of pressure drop calculation modules for a wire-wrapped LMR subassembly

International Nuclear Information System (INIS)

Kim, Young Gyun; Lim, Hyun Jin; Kim, Won Seok; Kim, Young Il

2000-06-01

Pressure drop calculation modules for a wire-wrapped LMR subassembly was been developed. This report summarizes present information on pressure drop calculation modules for inlet hole, lower part and upper part of a wire-wrapped LMR subassembly which was developed using simple formulas of sudden expansion and sudden contraction. A case calculation study was done using design data of a KALIMER driver fuel subassembly. And the total pressure drop in the driver fuel subassembly, except for the bundle part, was calculated as 0.13 MPa, which is in the reasonable pressure drop range. The developed modules will be integrated in the total subassembly pressure drop calculation code with further improvements

LMFBR subassembly response to local pressure loadings: an experimental approach

International Nuclear Information System (INIS)

Marciniak, T.J.; Ash, J.E.; Marchertas, A.H.; Cagliostro, D.J.

1975-01-01

An experimental program to determine the response of LMFBR-type subassemblies to local subassembly accidents caused by pressure loadings is described. Some results are presented and compared with computer calculations

Nuclear reactor fuel sub-assemblies

International Nuclear Information System (INIS)

Ford, J.; Bishop, J.F.W.

1981-01-01

An improved fuel sub-assembly for liquid metal cooled fast breeder nuclear reactors is described which facilitates dismantling operations for reprocessing purposes. The method of dismantling is described. (U.K.)

Evaluation of copper for divider subassembly in MCO Mark IA and Mark IV scrap fuel baskets

International Nuclear Information System (INIS)

Graves, C.E.

1997-01-01

The K Basin Spent Nuclear Fuel (SNF) Project Multi-Canister Overpack (MCO) subprojection eludes the design and fabrication of a canister that will be used to confine, contain, and maintain fuel in a critically safe array to enable its removal from the K Basins, vacuum drying, transport, staging, hot conditioning, and interim storage (Goldinann 1997). Each MCO consists of a shell, shield plug, fuel baskets (Mark IA or Mark IV), and other incidental equipment. The Mark IA intact and scrap fuel baskets are a safety class item for criticality control and components necessary for criticality control will be constructed from 304L stainless steel. It is proposed that a copper divider subassembly be used in both Mark IA and Mark IV scrap baskets to increase the safety basis margin during cold vacuum drying. The use of copper would increase the heat conducted away from hot areas in the baskets out to the wall of the MCO by both radiative and conductive heat transfer means. Thus copper subassembly will likely be a safety significant component of the scrap fuel baskets. This report examines the structural, cost and corrosion consequences associated with using a copper subassembly in the stainless steel MCO scrap fuel baskets

Thermal-hydraulic of partially blocked fuel subassembly with porous media

International Nuclear Information System (INIS)

Nagata, Takemitsu; Ohshima, Hiroyuki

2000-10-01

The analysis code for investigations of local subassembly phenomena, which has been recognized as an issue of local subassembly accidents, has been required and developed at JNC. It is desirable for the analysis code to be applicable to various blockage conditions and random position of the blockage formation and to evaluate conservatively on the safety assessment with high accuracy. In this study, for the purpose of verifying the application and issues of the subchannel analysis code ASFRE-IV which evaluates thermal hydraulic phenomena in the porous blockage regions, the ASFRE-IV validation analysis was carried out on the basis of the data of an experiment investigation on a local porous blockage in a fuel subassembly performed by Reactor Engineering Groop, O-arai Engineering Center, JNC. Calculational results indicated that ASFRE-IV could reproduce the coolant temperature profile in a fuel subassembly and the peak temperature in the local subchannel conservatively. (author)

Integrated intra-subassembly treatment in the SASSYS-1 LMR systems analysis code

Energy Technology Data Exchange (ETDEWEB)

Dunn, F.

1992-09-01

This report discusses a hot channel treatment which has been added to the SASSYS-1 LMR systems analysis code by providing for a multiple pin treatment of each of one or more subassemblies. This is an explicit calculation of intra-subassembly effects, not a hot-channel adjustment to a calculated average channel. Thus, the code can account for effects such as transient flow redistribution, both within a subassembly and among subassemblies. The code now provides a total integrated thermal hydraulic treatment including a multiple pin treatment within subassemblies, a multi-channel treatment of the whole core, and models for the primary coolant loops, the intermediate coolant loops, the steam generators, and the balance of plant. Currently the multiple-pin option is only implemented for single-phase calculations. It is not applicable after the onset of boiling or pin disruption. The new multiple pin treatment is being verified with detailed temperature data from instrumented subassemblies in EBR-II, both steady-state and transient, with special emphasis on passive safety tests such as SHRT-45. For the SHRT-45 test, excellent agreement is obtained between code predictions and experimental measurements of coolant temperatures.

Integrated intra-subassembly treatment in the SASSYS-1 LMR systems analysis code

Energy Technology Data Exchange (ETDEWEB)

Dunn, F.

1992-01-01

This report discusses a hot channel treatment which has been added to the SASSYS-1 LMR systems analysis code by providing for a multiple pin treatment of each of one or more subassemblies. This is an explicit calculation of intra-subassembly effects, not a hot-channel adjustment to a calculated average channel. Thus, the code can account for effects such as transient flow redistribution, both within a subassembly and among subassemblies. The code now provides a total integrated thermal hydraulic treatment including a multiple pin treatment within subassemblies, a multi-channel treatment of the whole core, and models for the primary coolant loops, the intermediate coolant loops, the steam generators, and the balance of plant. Currently the multiple-pin option is only implemented for single-phase calculations. It is not applicable after the onset of boiling or pin disruption. The new multiple pin treatment is being verified with detailed temperature data from instrumented subassemblies in EBR-II, both steady-state and transient, with special emphasis on passive safety tests such as SHRT-45. For the SHRT-45 test, excellent agreement is obtained between code predictions and experimental measurements of coolant temperatures.

STRUCTURE FOR SUB-ASSEMBLIES OF ELECTRONIC EQUIPMENT

Science.gov (United States)

Bell, P.R.; Harris, C.C.

1959-03-31

Sub-assemblies for electronic systems, particularly a unit which is self- contained and which may be adapted for quick application to and detachment from a chassis or panel, are discussed. The disclosed structure serves the dual purpose of a cover or enclosure for a subassembly comprising a base plate and also acts as a clamp for retaining the base plate in position on a chassis. The clamping action is provided by flexible fingers projecting from the side walls of the cover and extending through grooves in the base plate to engage with the opposite side of the chassis.

Coincidence measurements of FFTF breeder fuel subassemblies

International Nuclear Information System (INIS)

Eccleston, G.W.; Foley, J.E.; Krick, M.; Menlove, H.O.; Goris, P.; Ramalho, A.

1984-04-01

A prototype coincidence counter developed to assay fast breeder reactor fuel was used to measure four fast-flux test facility subassemblies at the Hanford Engineering Development Laboratory in Richland, Washington. Plutonium contents in the four subassemblies ranged between 7.4 and 9.7 kg with corresponding 240 Pu-effective contents between 0.9 and 1.2 kg. Large count rates were observed from the measurements, and plots of the data showed significant multiplication in the fuel. The measured data were corrected for deadtime and multiplication effects using established formulas. These corrections require accurate knowledge of the plutonium isotopics and 241 Am content in the fuel. Multiplication-corrected coincidence count rates agreed with the expected count rates based on spontaneous fission-neutron emission rates. These measurements indicate that breeder fuel subassemblies with 240 Pu-effective contents up to 1.2 kg can be nondestructively assayed using the shift-register electronics with the prototype counters. Measurements using the standard Los Alamos National Laboratory shift-register coincidence electronics unit can produce an assay value accurate to +-1% in 1000 s. The uncertainty results from counting statistics and deadtime-correction errors. 3 references, 8 figures, 8 tables

Reliability of wind turbine subassemblies

NARCIS (Netherlands)

Spinato, F.; Tavner, P.J.; Bussel, van G.J.W.; Koutoulakos, E.

2009-01-01

We have investigated the reliability of more than 6000 modern onshore wind turbines and their subassemblies in Denmark and Germany over 11 years and particularly changes in reliability of generators, gearboxes and converters in a subset of 650 turbines in Schleswig Holstein, Germany. We first start

The swelling behavior of Ti-stabilized austenitic steels used as structural materials of fissile subassemblies in Phenix

International Nuclear Information System (INIS)

Seran, J.L.; Touron, H.; Maillard, A.; Dubuisson, P.; Hugot, J.P.; Blanchard, P.; Pelletier, M.

1988-06-01

In this paper we analyse the main results obained on pressurized tubes, fissile pins and hexagonal cans, allowing us to characterize the swelling and irradiation creep resistance of Ti-Mod. austenitic steels, used as reference materials for the fast breeder subassembly. After having compared the global behavior of 316Ti and 15-15Ti steels irradiated as fissile pins we examine in more detail the leading variables acting on swelling and irradiation creep resistance of CW 316Ti clads and wrappers. The irradiation creep associated to the principal mechanical stresses (sodium pressure for the wrapper, fission gas pressure for the clad) explain the plastic deformation observed on the wrappers not on the clads. Fissile pins swell more and the scatter of the results is larger than for wrappers or samples. It does not seem possible to invoque flux or primary stress differences to explain this fact. On the opposite the thermal gradient in the thickness of the components appears to be a significant parameter. In fissile pins it gives rise to a swelling gradient observed by electron microscopy that must be taken into account when comparing to the wrapper. As compared to CW 316Ti, CW 15-15Ti is an important improvement since its incubation dose for swelling is far beyond 100 dpa. Further more since it swelling temperature dependence does not seem to be as important as for 316Ti, it should be less sensitive to the effect of thermal gradients

Nondestructive assay of subassemblies of various spent or fresh fuels by active neutron interrogation

International Nuclear Information System (INIS)

Ragan, G.L.; Ricker, C.W.; Chiles, M.M.; Ingersoll, D.T.; Slaughter, G.G.

1979-01-01

Recent studies show that subassemblies containing various spent fuels could be assayed rapidly and accurately by a nondestructive assay system using active neutron interrogation and prompt-neutron detection. Subassembly penetration is achieved by 24-keV (Sb--Be) interrogation neutrons; the spent-fuel neutron background is overridden by using strong interrogating sources and prompt-neutron signals, and background gammas are absorbed by lead. Experiments have demonstrated the potential for assaying with better than 5% accuracy, three spent plutonium-fueled subassemblies per hour. Calculations, validated by experiments, predict even better performance for fresh or uranium-fueled subassemblies; several performance estimates are given

Nuclear fuel sub-assemblies

International Nuclear Information System (INIS)

Dodd, J.A.; Butterfield, C.E.; Waite, E.

1979-01-01

A fast reactor fuel sub-assembly has honeycomb grids for laterally supporting the fuel pins. The grids are of two series and are arranged alternately along the bundle. The grids of a first series provide a discrete cell for each pin but the grids of the second series have a peripheral group of cells only. The grids of the second series provide intermediate support of the edge pins to restrain bow. (author)

Studies to single subassembly flow monitoring with a complete 7 element array under sodium

International Nuclear Information System (INIS)

Hess, B.; Ruppert, E.; Stehle, H.; Vinzens, K.

1975-01-01

A core restraint system in a fast reactor serves to limit fuel element movement leading to reactivity changes and misalignment of control rod drives and instrumentation. To guarantee proper control rod function the upper ring of the passive restraint system for the SNR-300 should keep subassembly displacement below 20 mm, whereas a free bowing up to 25 mm does not impair subassembly handling. With respect to single subassembly instrumentation the influences of subassembly displacement on temperature and flow monitoring were not exactly known. As part of the SNR-300 R and D programme a complete clamped array, consisting of 4 full size fuel elements and 3 blanket elements was tested for more than 4000 hours at 600 0 C in the AKB sodium loop at Interatom, Bensberg. The test was split into two phases and the total cluster was prestrained in the second phase to simulate 15 mm subassembly displacement at the level of the upper pads. Although this test was mainly considered as an endurance test to demonstrate the integrity of prestrained core elements, effort were made to study the feasibility of single subassembly flow monitoring with this full size model of a core section. (Auth.)

Fuel and fuel pin behaviour in a high burnup fast breeder fuel subassembly: Results of destructive post-irradiation examinations of the KNK II/1 fuel subassembly NY-205

International Nuclear Information System (INIS)

Patzer, G.

1991-05-01

The report gives a summarizing overview of the design characteristics, of the irradiation history and of the results of the destructive post-irradiation examinations of the fuel pins of the high-burnup fuel subassembly NY-205 of the KNK II first core. This element was operated for about 10 years and reached a maximum local burnup of 175 MWd/kg(HM) and a maximum neutron dose of 67 dpa-NRT. The main design data of this subassembly agree with those of the SNR 300 Mark-Ia, and it reached more than twice of the burnup and a similar neutron dose as foreseen for the SNR 300 fuel subassemblies [de

Nuclear reactor fuel element sub-assemblies

International Nuclear Information System (INIS)

Hill, G.D.; Trevalion, P.A.

1977-01-01

A fuel element sub-assembly for a liquid metal cooled fast reactor is described. It comprises a bundle of fuel pins enclosed by a tubular wrapper having a lower end journal for plugging into an upper aperture in a core supporting structure and a spike bar with an articulated bush for engaging a lower aperture in the core supporting structure. The articulated bush is retained on a spherical end portion of the spike bar by a pair of parallel retaining pins arranged transversely and disposed one each side of the spike bar. The pins are tubular and collapsible at a predetermined loading to enable the spherical end portion to pass between them. The articulated bush has an internal groove for engagement by a lifting grab, this groove being formed in a bore for receiving the spherical end portion of the spike bar. The construction lessens liability to rattling of the fuel element sub-assemblies and aids removal for replacement. (U.K.)

Response of subassembly model with internals

International Nuclear Information System (INIS)

Kennedy, J.M.; Belytschko, T.

1977-01-01

For the purpose of predicting the structural response in such accident environments, a program STRAW has been developed. This is a finite element program which can treat the structure-fluid system consisting of the coolant and the subassembly walls. Both material nonlinearities due to elastic-plastic response and geometric nonlinearities due to large displacements can be treated. The energy source can be represented either by a pressure-time history or an equation of state. Because of the lack of any simplifying symmetry in the geometry of the subassembly the program uses a quasi-three dimensional model. The cross section of the accident hexcan and the adjacent hexcan are modelled by a two-dimensional finite element mesh which represents the hexcan walls by flexural element and the internals by two-dimensional continuum elements. This mesh is coupled to a series of one-dimensional elements which represent the axial flow of the coolant and the longitudinal stiffness of the fuel pins and hexcan. The latter is of importance in the adjacent hexcan, for its lateral displacement is resisted entirely by this flexural behavior and its inertia. The adequacy of such quasi-three dimensional models has been examined by comparing the STRAW results against almost complete three-dimensonal analysis performed with the REXCAT program. In this program, the accident hexcan is represented in a true three-dimensional sense by plate-shell elements, whereas the internals are represented as axisymmetric. These comparisons indicate that the quasi-three-dimensional approach employed in STRAW is valid for a large range of pressure time histories; the fidelity of this model suffers primarily when pressure reaches a peak over a very short time, such as 5-10 microseconds

Dynamic response of single hexagonal LMFBR core subassembly wrappers

Energy Technology Data Exchange (ETDEWEB)

Ash, J. E.; Marciniak, T. J.; (Argonne National Lab., IL (United States))

1977-07-01

To analyze the dynamic structural response of the LMFBR core subassembly hexagonal wrappers to postulated local energy releases and the sensitivity of the response to variations in both the pressure loading and the material properties of the stainless steel, a finite-element computer code STRAW has been developed. A series of experiments was performed to study the effects of variations in material properties. The amount of coldworking to which the Type 316 stainless steel is subjected has a strong influence upon the ductility and the elastic yield point. The usual fabrication process produced a nominally 20% coldworking with a yield point of about 680 MPa. By designing a special set of dies for the drawing process, a very low ductility hexcan was produced for which the yield point was raised to 820 MPa. Conversely, the yield point was lowered to 170 MPa by a solution annealing process producing a highly ductile test hexcan. A metallurgical study was conducted to find a representative brittle simulant material for the irradiated end-of-life steel properties. An aging treatment for Type 446 stainless steel was developed which reproduced the expected tensile-flow behavior of the in-pile subassembly. Further study is underway to investigate the fracture properties of the simulant material. The pressure pulses were generated by the controlled expansion of high-pressure detonation poducts from low-density explosives detonated inside a vented steel cannister. The orifice configuration of the cannister and the charge mixture ratio were designed to produce two specified pulse shapes. A charge containing 37,7 g PETN mixed with 35 wt % inert, hollow-glass microballoons developed a pressure pulse peak of 9.5 MPa at 1.0 ms. Increasing the PETN to 41 g resulted in a 14.6 MPa peak pressure, and increasing the explosive concentration to 90 wt % in the mixture increased the burning rate and the pulse risetime, so that the peak occurred at 0.6 ms.

Time constants and feedback transfer functions of EBR-II subassembly types

International Nuclear Information System (INIS)

Grimm, K.N.; Meneghetti, D.

1986-01-01

Time constants, feedback reactivity transfer functions and power coefficients are calculated for stereotypical subassemblies in the EBR-II reactor. These quantities are calculated from nodal reactivities obtained from a reactor kinetic code analysis for a step change in power. Due to the multiplicity of eigenvalues, there are several time constants for each nodal position in a subassembly. Compared with these calculated values are analytically derived values for the initial node of a given channel

Time constants and feedback transfer functions of EBR-II subassembly types

International Nuclear Information System (INIS)

Grimm, K.N.; Meneghetti, D.

1987-01-01

Time constants, feedback reactivity transfer functions and power coefficients are calculated for stereotypical subassemblies in the EBR-II reactor. These quantities are calculated from nodal reactivities obtained from a reactor kinetic code analysis for a step change in power. Due to the multiplicity of eigenvalues, there are several time constants for each nodal position in a subassembly. Compared with these calculated values are analytically derived values for the initial node of a given channel. (author)

Fuel or irradiation subassembly

International Nuclear Information System (INIS)

Seim, O.S.; Hutter, E.

1975-01-01

A subassembly for use in a nuclear reactor is described which incorporates a loose bundle of fuel or irradiation pins enclosed within an inner tube which in turn is enclosed within an outer coolant tube and includes a locking comb consisting of a head extending through one side of the inner sleeve and a plurality of teeth which extend through the other side of the inner sleeve while engaging annular undercut portions in the bottom portion of the fuel or irradiation pins to prevent movement of the pins

Proof tests of irradiated and unirradiated EBR-II subassembly ducts

International Nuclear Information System (INIS)

Ruther, W.E.; Chopra, P.S.; Lambert, J.D.B.

1977-01-01

A series of dynamic pressure tests have been conducted within EBR-II subassembly ducts. The tests were designed to simulate bursting of a driver-fuel element in a cluster of such elements at their burnup limit during off-normal conditions in EBR-II. The major objective of the tests was to assure that such failure, which might cause rapid release of stored fission gas, would not deform or otherwise damage subassembly ducts in a way that would hinder movement of a control rod. The test results are described

Possible new basis for fast reactor subassembly instrumentation

Energy Technology Data Exchange (ETDEWEB)

Edwards, A G

1977-03-01

This is a digest of a paper presented to the Risley Engineering Society. The theme is a speculation that the core instrumentation problem for a liquid metal fast breeder reactor might be transformed by developments in the realm of infrared television and in pattern recognition by computer. There is a possible need to measure coolant flow and cooled exit temperature for each subassembly, with familiar fail-to-safety characteristics. Present methods use electrical devices, for example thermocouples, but this gives rise to cabling problems. It might be possible, however, to instal at the top of each subassembly a mechanical device that gives a direct indication of temperature and flow visible to an infrared television camera. Signal transmission by cable would then be replaced by direct observation. A possible arrangement for such a system is described and is shown in schematic form. It includes pattern recognition by computer. It may also be possible to infer coolant temperature directly from the characteristics of the infrared radiation emitted by a thin stainless steel sheet in contact with the sodium, and an arrangement for this is shown. The type of pattern produced for on-line interpretation by computer is also shown. It is thought that this new approach to the problem of subassembly instrumentation is sufficiently attractive to justify a close study of the problems involved.

Fuel sub-assembly

International Nuclear Information System (INIS)

Jolly, R.

1988-01-01

A nuclear fuel sub-assembly includes a hexagonal bundle of parallel, spaced apart fuel pins coupled at one end to an end-holding grid comprising a number of transverse spaced apart rails to each of which is connected a series of pin-receiving cells which render the pins axially captive with the rails. The series of cells are defined by a pair of metal strips each of which has a series of pocket formations such that when the pocket formations are in registry they define cylindrical shaped cells provided with internal projections which engage annular recesses in the end caps of the fuel pins to effect axial constraint of the pins. (author)

Desain Sistem Pendeteksi untuk Citra Base Sub-assembly dengan Algoritma Backpropagation

Directory of Open Access Journals (Sweden)

Kasdianto Kasdianto

2017-04-01

Full Text Available Object identification technique using machine vision has been implemented in industrial of electronic manufacturers for years. This technique is commonly used for reject detection (for disqualified product based on existing standard or defect detection. This research aims to build a reject detector of sub-assembly condition which is differed by two conditions that are missing screw and wrong position screw using neural network backpropagation. The image taken using camera will be converted into grayscale before it is processed in backpropagation methods to generate a weight value. The experiment result shows that the network architecture with two layers has the most excellent accuracy level. Using learning rate of 0.5, target error 0.015%, and the number of node 1 of 100 and node 2 of 50, the successive rate for sub-assembly detection in right condition reached 99.02% while no error occurs in detecting the wrong condition of Sub-assembly (missing screw and wrong position screw.

System for nondestructive assay of spent fuel subassemblies: comparison of calculations and measurements

International Nuclear Information System (INIS)

Ragan, G.L; Ricker, C.W.; Chiles, M.M.; Ingersoll, D.T.; Slaughter, G.G.; Williams, L.R.

1979-01-01

A nondestructive assay system was developed for determining the total fissile content of spent fuel subassemblies at the head end of a reprocessing plant. The system can perform an assay in 20 min with an uncertainty of <5%. Antimony-beryllium neutrons interrogate the subassemblies, and proton recoil counters detect the resulting fission neutrons. Pulse-height discrimination differentiates between the low-energy interrogation neutrons and the higher-energy fission neutrons. Calculated and measured results were compared for (1) interrogation-neutron penetrability, (2) fission-neutron detectability, (3) radial variation of assay sensitivity, (4) axial variation of assay sensitivity, and (5) the variation of detector count rate as a function of the number of fuel rods in a special 61-rod, LMFBR-type subassembly

High 240Pu FTR/EMC experiments and analysis: Carbide fuel and UO2 blanket subassembly worths

International Nuclear Information System (INIS)

Ombrellaro, P.A.

1977-06-01

Carbide-plutonium fuel and UO 2 blanket subassembly worth measurements performed at ANL in the EMC/LWR were analyzed. Composition exchange worth calculations were performed for: (a) the replacement of high- 240 Pu fuel composition for low- 240 Pu fuel composition and carbide-plutonium fuel composition, successively, in the center subassembly of the core; (b) the replacement of low- 240 Pu fuel composition for carbide--plutonium fuel composition in one outer driver subassembly; and (c) the replacement of the radial reflector composition with UO 2 blanket composition in one subassembly of the radial reflector. The composition exchange worth calculations were performed in two-dimensional x,y geometry, using diffusion theory and perturbation theory. Each method produces about the same calculated-to-experimental bias factors

Automated cleaning of electronic components

International Nuclear Information System (INIS)

Drotning, W.; Meirans, L.; Wapman, W.; Hwang, Y.; Koenig, L.; Petterson, B.

1994-01-01

Environmental and operator safety concerns are leading to the elimination of trichloroethylene and chlorofluorocarbon solvents in cleaning processes that remove rosin flux, organic and inorganic contamination, and particulates from electronic components. Present processes depend heavily on these solvents for manual spray cleaning of small components and subassemblies. Use of alternative solvent systems can lead to longer processing times and reduced quality. Automated spray cleaning can improve the quality of the cleaning process, thus enabling the productive use of environmentally conscious materials, while minimizing personnel exposure to hazardous materials. We describe the development of a prototype robotic system for cleaning electronic components in a spray cleaning workcell. An important feature of the prototype system is the capability to generate the robot paths and motions automatically from the CAD models of the part to be cleaned, and to embed cleaning process knowledge into the automatically programmed operations

Development of thermohydraulic codes for modeling liquid metal boiling in LMR fuel subassemblies

International Nuclear Information System (INIS)

Sorokin, G.A.; Avdeev, E.F.; Zhukov, A.V.; Bogoslovskaya, G.P.; Sorokin, A.P.

2000-01-01

An investigation into the reactor core accident cooling, which are associated with the power grow up or switch off circulation pumps in the event of the protective equipment comes into action, results in the problem of liquid metal boiling heat transfer. Considerable study has been given over the last 30 years to alkaline metal boiling including researches of heat transfer, boiling patterns, hydraulic resistance, crisis of heat transfer, initial heating up, boiling onset and instability of boiling. The results of these investigations have shown that the process of liquid metal boiling has substantial features in comparison with water boiling. Mathematical modeling of two phase flows in fast reactor fuel subassemblies have been developed intensively. Significant success has been achieved in formulation of two phase flow through the pin bundle and in their numerical realization. Currently a set of codes for thermohydraulic analysis of two phase flows in fast reactor subassembly have been developed with 3D macrotransfer governing equations. These codes are used for analysis of boiling onset and liquid metals boiling in fuel subassemblies during loss-of-coolant accidents, of warming up of reactor core, of blockage of some part of flow cross section in fuel subassembly. (author)

A possible new basis for fast reactor subassembly instrumentation

International Nuclear Information System (INIS)

Edwards, A.G.

1977-01-01

This is a digest of a paper presented to the Risley Engineering Society. The theme is a speculation that the core instrumentation problem for a liquid metal fast breeder reactor might be transformed by developments in the realm of infrared television and in pattern recognition by computer. There is a possible need to measure coolant flow and cooled exit temperature for each subassembly, with familiar fail-to-safety characteristics. Present methods use electrical devices, for example thermocouples, but this gives rise to cabling problems. It might be possible, however, to instal at the top of each subassembly a mechanical device that gives a direct indication of temperature and flow visible to an infrared television camera. Signal transmission by cable would then be replaced by direct observation. A possible arrangement for such a system is described and is shown in schematic form. It includes pattern recognition by computer. It may also be possible to infer coolant temperature directly from the characteristics of the infrared radiation emitted by a thin stainless steel sheet in contact with the sodium, and an arrangement for this is shown. The type of pattern produced for on-line interpretation by computer is also shown. It is thought that this new approach to the problem of subassembly instrumentation is sufficiently attractive to justify a close study of the problems involved. (U.K.)

Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

International Nuclear Information System (INIS)

Kamide, H.; Ieda, Y.; Toda, S.; Isozaki, T.; Sugawara, S.

1993-01-01

Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor core during natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

Energy Technology Data Exchange (ETDEWEB)

Kamide, H; Ieda, Y; Toda, S; Isozaki, T; Sugawara, S [Reactor Engineering Section, O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation, Narita, O-arai, Ibaraki-ken (Japan)

1993-02-01

Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor coreduring natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

Recommended connections between the main ionizing radiation sensors and their electronic sub-assemblies

International Nuclear Information System (INIS)

Lefevre, Roger; Roquefort, Henri

1970-02-01

The authors report the study of several typical and simple connections which are present between an ionizing radiation detector and the electronic sub-assembly, and can be adequate in most of the cases. They also study recommended outputs of the different types of detectors and their possible connections with electronic functional elements. Thus, they address connections for general use, detector outputs, inputs and outputs of amplifiers and of sub-assemblies, amplifier inputs, commonly used connectors

Universal Fast Breeder Reactor Subassembly Counter manual

International Nuclear Information System (INIS)

Menlove, H.O.; Eccleston, G.W.; Swansen, J.E.; Goris, P.; Abedin-Zadeh, R.; Ramalho, A.

1984-08-01

A neutron coincidence counter has been designed for the measurement of fast breeder reactor fuel assemblies. This assay system can accommodate the full range of geometries and masses found in fast breeder subassemblies under IAEA safeguards. The system's high-performance capability accommodates high plutonium loadings of up to 16 kg. This manual describes the system and its operation and gives performance and calibration parameters for typical applications

Dynamic structural response of reactor-core subassemblies (hexcans) due to accident overpressurization

International Nuclear Information System (INIS)

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

1993-01-01

This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall. (author)

Dynamic structural response of reactor-core subassemblies (hexcans) due to accident overpressurization

International Nuclear Information System (INIS)

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

1993-01-01

This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall

Calculation of the mechanical equilibrium in a lattice of deformed hexagonal subassemblies

International Nuclear Information System (INIS)

Bernard, A.

1979-01-01

Stainless steel swelling and irradiation creep in the hexagonal wrappers of fast breeder cores induce deformations (mostly bowing), hence mutual interaction (displacements, forces and stresses, which must be calculated). The HARMONIE code was developed to meet these requirements. In this three dimensional code, one minimizes the elastic potential bending energy (quadratic form), with given linear conditions (no overlapping between adjacent subassemblies). The convergence of this function is obtained through a numerical method (parallel gradient). The free bowing of the subassemblies are given as input datas; the output gives the equilibrium displacements and forces while stresses are calculated in a classical manner

Study on mixed convective flow penetration into subassembly from reactor hot plenum in FBRs

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, J.; Ohshima, H.; Kamide, H.; Ieda, Y. [Power Reactor and Nuclear Fuel Development Corporation, Ibaraki (Japan)

1995-09-01

Fundamental experiments using water were carried out in order to reveal the phenomenon of mixed convective flow penetration into subassemblies from a reactor`s upper plenum of fast breeder reactors. This phenomenon appears under a certain natural circulation conditions during the operation of the direct reactor auxiliary cooling system for decay heat removal and might influence the natural circulation head which determines the core flow rate and therefore affects the core coolability. In the experiment, a simplified model which simulates an upper plenum and a subassembly was used and the ultrasonic velocity profile monitor as well as thermocouples were applied for the simultaneous measurement of velocity and temperature distributions in the subassembly. From the measured data, empirical equations related to the penetration flow onset condition and the penetration depth were obtained using relevant parameters which were derived from dimensional analysis.

Universal Fast Breeder Reactor Subassembly Counter manual

Energy Technology Data Exchange (ETDEWEB)

Menlove, H.O.; Eccleston, G.W.; Swansen, J.E.; Goris, P.; Abedin-Zadeh, R.; Ramalho, A.

1984-08-01

A neutron coincidence counter has been designed for the measurement of fast breeder reactor fuel assemblies. This assay system can accommodate the full range of geometries and masses found in fast breeder subassemblies under IAEA safeguards. The system's high-performance capability accommodates high plutonium loadings of up to 16 kg. This manual describes the system and its operation and gives performance and calibration parameters for typical applications.

Time constants and feedback transfer functions of EBR-II [Experimental Breeder Reactor] subassembly types

International Nuclear Information System (INIS)

Grimm, K.N.; Meneghetti, D.

1986-09-01

Time constants, feedback reactivity transfer functions and power coefficients are calculated for stereotypical subassemblies in the EBR-II reactor. These quantities are calculated from nodal reactivities obtained from a reactor kinetic code analysis for a step change in power. Due to the multiplicity of eigenvalues, there are several time constants for each nodal position in a subassembly. Compared with these calculated values are analytically derived values for the initial node of a given channel

Assessment of MATRA-LMR-FB with the SHRT-17 Core Subassembly Data

Energy Technology Data Exchange (ETDEWEB)

Chang, Won-Pyo; Yoo, Jin Yoo; Lee, Seung Won; Seong, Seung Hwan; Ahn, Sang June; Choi, Chi-Woong; Lee, Kwi Lim; Jeong, Jae-Ho; Jeong, Taekyeong; Ha, Kwi-Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

Since the MATRA-LMRFB code is scheduled to be applied to a partial flow blockage analysis in a PGSFR (Prototype Generation IV Sodium-cooled Fast Reactor) subassembly, code verification is an essential part for the design review. Most of its verification efforts have been devoted to local sub-channel blockages, and thus the predictions were compared to those of other codes as well as experimental data. Verifications using pin bundles with a wire-wrap spacer had to be concentrated to 19-pin bundles, because available experimental data for such a bundle were relatively affluent in world-wide literatures. Therefore, more assessments with diverse pin numbers are necessary for MATRA-LMR-FB to be a more reliable code. Thus far, MATRA-LMR-FB has been applied to a 37-pin subassembly with wire-wrap spacers at most. In this regard, the present comparative study using data produced from the SHRT-17 which was carried out in a 61-pin test subassembly (XX09) placed in the EBR-II (Experimental Breeder Reactor II) core will be a meaningful demonstration for its extensive applicability. The power operation of the EBR-II was begun by ANL (Argonne National Lab.) in 1964 and the SHRT program was carried out in EBR-II between 1984 and 1986 in order to provide not only test data for validation of the computer codes but also demonstration of a passive reactor shutdown and decay heat removal in response of the protected and unprotected transients. The EBR-II SHRT-17 test data were used to demonstrate the prediction capability of MATRA-LMRFB on a radial distribution of the subassembly outlet temperatures during the steady state.

Assessment of MATRA-LMR-FB with the SHRT-17 Core Subassembly Data

International Nuclear Information System (INIS)

Chang, Won-Pyo; Yoo, Jin Yoo; Lee, Seung Won; Seong, Seung Hwan; Ahn, Sang June; Choi, Chi-Woong; Lee, Kwi Lim; Jeong, Jae-Ho; Jeong, Taekyeong; Ha, Kwi-Seok

2015-01-01

Since the MATRA-LMRFB code is scheduled to be applied to a partial flow blockage analysis in a PGSFR (Prototype Generation IV Sodium-cooled Fast Reactor) subassembly, code verification is an essential part for the design review. Most of its verification efforts have been devoted to local sub-channel blockages, and thus the predictions were compared to those of other codes as well as experimental data. Verifications using pin bundles with a wire-wrap spacer had to be concentrated to 19-pin bundles, because available experimental data for such a bundle were relatively affluent in world-wide literatures. Therefore, more assessments with diverse pin numbers are necessary for MATRA-LMR-FB to be a more reliable code. Thus far, MATRA-LMR-FB has been applied to a 37-pin subassembly with wire-wrap spacers at most. In this regard, the present comparative study using data produced from the SHRT-17 which was carried out in a 61-pin test subassembly (XX09) placed in the EBR-II (Experimental Breeder Reactor II) core will be a meaningful demonstration for its extensive applicability. The power operation of the EBR-II was begun by ANL (Argonne National Lab.) in 1964 and the SHRT program was carried out in EBR-II between 1984 and 1986 in order to provide not only test data for validation of the computer codes but also demonstration of a passive reactor shutdown and decay heat removal in response of the protected and unprotected transients. The EBR-II SHRT-17 test data were used to demonstrate the prediction capability of MATRA-LMRFB on a radial distribution of the subassembly outlet temperatures during the steady state

Thermal hydraulic behavior of sub-assembly local blockage in China experiment fast reactor

International Nuclear Information System (INIS)

Yang Zhimin

2000-01-01

The geometrical parameter ratio of pitch to diameter of China Experiment Fast Reactor (CEFR) subassembly is 1,167. To address the thermal hydraulic behavior of subassembly local blockage which may be caused by deformation of cladding due to severe swelling and thermal stresses and by space swirl loosening etc., the porous numerical model and SIMPLE-P code used to solve Navier-Stokes and energy equations in porous medium was developed, and the bundle experiment with 19 pins with 24 subchannels blocked in the sodium coolant was carried on in China Institute of Atomic Energy (CIAE). The comparison of code predictions against experiments (including non-blockage and ten blockage conditions) seems well. The thermal hydraulic behavior of fuel subassembly with 61 fuel pins blockage of CEFR is calculated with SIMPLE-P code. The results indicate that the maximum temperature is 815 deg. C when the blockage area is about 37% (54 central subchannels are blocked). In this case the cladding won't be damaged and no sodium coolant boiling takes place. (author)

Experimental confirmation of the design to minimize vibration and wear in 61-pin wire-spaced EBR-II subassemblies

International Nuclear Information System (INIS)

Fukuda, S.K.

1978-05-01

Examinations of HEDL 61-pin subassemblies comprised of 5.84 mm (0.230) inch diameter mixed-oxide fuel pins with 1.02 mm (0.040'') diameter spacer wire (PNL-9, -10, -11, HEDL-N-E, -N-F), showed severe cladding and spacer wire wear after irradiation in EBR-II. A comparison of a large number of design, fabrication, and irradiation parameters for all of the HEDL subassemblies indicated that the porosity per ring of fuel pins correlated significantly with the occurrence of wear on the fuel pins. The porosity per ring is the clearance between the flat-to-flat pin bundle dimension and the inner hex can dimension divided by the number of hexagonal fuel pin rings in the subassembly. The porosity per ring for PNL-9, -10, -11 and HEDL-N-E was 0.15 mm/ring (6 mils/ring) and 0.18 mm/ring (7 mils/ring) for the HEDL-N-F subassembly. Since the original FTR subassembly design had a porosity/ring spread of 0.04 mm/ring to 0.16 mm/ring (1.67 to 6.11 mils/ring) an additional series of irradiation tests was conducted to confirm that a tighter fuel pin bundle would eliminate the wear

Development of materials and manufacturing technologies for Indian fast reactor programme

International Nuclear Information System (INIS)

Raj, Baldev; Jayakumar, T.; Bhaduri, A.K.; Mandal, Sumantra

2010-01-01

Fast Breeder Reactors (FBRs) are vital towards meeting security and sustainability of energy for the growing economy of India. The development of FBRs necessitates extensive research and development in domains of materials and manufacturing technologies in association with a wide spectrum of disciplines and their inter-twining to meet the challenging technology. The paper highlight the work and the approaches adopted for the successful deployment of materials, manufacturing and inspection technologies for the in-core and structural components of current and future Indian Fast Breeder Reactor Programme. Indigenous development of in-core materials viz. Titanium modified austenitic stainless steel (Alloy D9) and its variants, ferritic/martensitic oxide-dispersion strengthened (ODS) steels as well as structural materials viz. 316L(N) stainless steel and modified 9Cr-1Mo have been achieved through synergistic interactions between Indira Gandhi Centre for Atomic Research (IGCAR), education and research institutes and industries. Robust manufacturing technology has been established for forming and joining of various components of 500 MWe Prototype Fast Breeder Reactor (PFBR) through 'science-based technology' approach. To achieve the strict quality standards of formed parts in terms of geometrical tolerances, residual stresses and microstructural defects, FEM-based modelling and experimental validation was carried out for estimation of spring-back during forming of multiple curvature thick plantes. Optimization of grain boundary character distribution in Alloy D9 was carried out by adopting the grain boundary engineering approach to reduce radiation induced segregation. Extensive welding is involved in the fabrication of reactor vessels, piping, steam generators, fuel sub-assemblies etc. Activated Tungsten Inert Gas Welding process along with activated flux developed at IGCAR has been successfully used in fabrication of dummy fuel subassemblies (DFSA) required for testing

Development of materials and manufacturing technologies for Indian fast reactor programme

Energy Technology Data Exchange (ETDEWEB)

Raj, Baldev; Jayakumar, T.; Bhaduri, A.K.; Mandal, Sumantra [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

2010-07-01

Fast Breeder Reactors (FBRs) are vital towards meeting security and sustainability of energy for the growing economy of India. The development of FBRs necessitates extensive research and development in domains of materials and manufacturing technologies in association with a wide spectrum of disciplines and their inter-twining to meet the challenging technology. The paper highlight the work and the approaches adopted for the successful deployment of materials, manufacturing and inspection technologies for the in-core and structural components of current and future Indian Fast Breeder Reactor Programme. Indigenous development of in-core materials viz. Titanium modified austenitic stainless steel (Alloy D9) and its variants, ferritic/martensitic oxide-dispersion strengthened (ODS) steels as well as structural materials viz. 316L(N) stainless steel and modified 9Cr-1Mo have been achieved through synergistic interactions between Indira Gandhi Centre for Atomic Research (IGCAR), education and research institutes and industries. Robust manufacturing technology has been established for forming and joining of various components of 500 MWe Prototype Fast Breeder Reactor (PFBR) through 'science-based technology' approach. To achieve the strict quality standards of formed parts in terms of geometrical tolerances, residual stresses and microstructural defects, FEM-based modelling and experimental validation was carried out for estimation of spring-back during forming of multiple curvature thick plantes. Optimization of grain boundary character distribution in Alloy D9 was carried out by adopting the grain boundary engineering approach to reduce radiation induced segregation. Extensive welding is involved in the fabrication of reactor vessels, piping, steam generators, fuel sub-assemblies etc. Activated Tungsten Inert Gas Welding process along with activated flux developed at IGCAR has been successfully used in fabrication of dummy fuel subassemblies (DFSA) required

Structural response of reactor-core hexcan subassemblies subjected to dynamic overpressurization under accident conditions

International Nuclear Information System (INIS)

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

1993-01-01

This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall

Analytical work on local faults in LMFBR subassembly

International Nuclear Information System (INIS)

Yoshikawa, H.; Miyaguchi, K.; Hirata, N.; Kasahara, F.

1979-01-01

Analytical codes have been developed for evaluating various severe but highly unlikely events of local faults in the LMFBR subassembly (S/A). These include: (1) local flow blockage, (2) two-phase thermohydraulics under fission gas release, and (3) inter-S/A failure propagation. A simple inter-S/A thermal failure propagation analysis code, FUMES, is described that allows an easy parametric study of propagation potential of fuel fog in a S/A. 7 refs

Nuclear sub-assembly for liquid metal fast breeder reactor

International Nuclear Information System (INIS)

1978-01-01

The description is given of a nuclear sub-assembly comprising several spaced out fuel pins in a tubular shroud, the characteristic being that the section of the shroud forms a closed figure with six main straight sides in hexagonal shape, the main sides being joined by subsidiary sides which are either straight or convex towards the centre of the figure [fr

Prediction of the pressure-time history due to fuel-sodium interaction in a subassembly

International Nuclear Information System (INIS)

Jacobs, H.

1975-01-01

A local cooling disturbance may lead to complete voiding of a subassembly and melt down of the fuel pins. Thus molten fuel may be accumulated and mixed with liquid sodium returning accidentally into the subassembly. The resulting fuel-sodium interaction (FSI) produces a pressure load on the surrounding core structures. It is necessary to prove that the corresponding core deformation neither initiates a nuclear excursion nor renders the shut down system inoperable. This requires the knowledge of the initiating FSI pressure time history. In this paper a theoretical pressure time history is presented which differs completely from all calculations known so far. (Auth.)

Identifying subassemblies by ultrasound to prevent fuel handling error in sodium fast reactors: First test performed in water

International Nuclear Information System (INIS)

Paumel, Kevin; Lhuillier, Christian

2015-01-01

Identifying subassemblies by ultrasound is a method that is being considered to prevent handling errors in sodium fast reactors. It is based on the reading of a code (aligned notches) engraved on the subassembly head by an emitting/receiving ultrasonic sensor. This reading is carried out in sodium with high temperature transducers. The resulting one-dimensional C-scan can be likened to a binary code expressing the subassembly type and number. The first test performed in water investigated two parameters: width and depth of the notches. The code remained legible for notches as thin as 1.6 mm wide. The impact of the depth seems minor in the range under investigation. (authors)

Direct current linear measurement sub-assembly data and test methods. Nuclear electronic equipment for control and monitoring panel

International Nuclear Information System (INIS)

1977-12-01

The M.C.H./M.E.N.T.3 document is concerned with sub-assemblies intended for measuring on a linear scale the neutron fluence rate or radiation dose rate when connected with nuclear detectors working in current. The symbols used are described. Some definitions and a bibliography are given. The main characteristics of direct current linear measurement sub-assemblies are then described together with corresponding test methods. This type of instrument indicates on a linear scale the level of a direct current applied to its input. The document reviews linear sub-assemblies for general purpose applications, difference amplifiers for monitoring, and averaging amplifiers. The document is intended for electronics manufacturers, designers, persons participating in acceptance trials and plant operators [fr

Development of multi-dimensional thermal-hydraulic modeling using mixing factors for wire wrapped fuel pin bundles in fast reactors. Validation through a sodium experiment of 169-pin fuel subassembly

International Nuclear Information System (INIS)

Nishimura, M.; Kamide, H.; Miyake, Y.

1997-04-01

Temperature distributions in fuel subassemblies of fast reactors interactively affect heat transfer from center to outer region of the core (inter-subassembly heat transfer) and cooling capability of an inter-wrapper flow, as well as maximum cladding temperature. The prediction of temperature distribution in the subassembly is, therefore one of the important issues for the reactor safety assessment. Mixing factors were applied to multi-dimensional thermal-hydraulic code AQUA to enhance the predictive capability of simulating maximum cladding temperature in the fuel subassemblies. In the previous studies, this analytical method had been validated through the calculations of the sodium experiments using driver subassembly test rig PLANDTL-DHX with 37-pin bundle and blanket subassembly test rig CCTL-CFR with 61-pin bundle. The error of the analyses were comparable to the error of instrumentation's. Thus the modeling was capable of predicting thermal-hydraulic field in the middle scale subassemblies. Before the application to large scale real subassemblies with more than 217 pins, accuracy of the analytical method have to be inspected through calculations of sodium tests in a large scale pin bundle. Therefore, computations were performed on sodium experiments in the relatively large 169-pin subassembly which had heater pins sparsely within the bundle. The analysis succeeded to predict the experimental temperature distributions. The errors of temperature rise from inlet to maximum values were reduced to half magnitudes by using mixing factors, compared to those of analyses without mixing factors. Thus the modeling is capable of predicting the large scale real subassemblies. (author)

Finite element analysis of irradiation-induced dilation of the fuel subassembly duct in LMFBR

International Nuclear Information System (INIS)

Gao Fuhai; Fu Hao; Li Nan; Yang Kongli; Wang Mingzhen

2013-01-01

Background: The calculation of irradiation-induced dilation of the fuel subassembly duct in LMFBR is important for fast reactor core design.. Purpose: To investigate how to calculate the dilation by using finite element method (FEM). Methods: First, irradiation-induced creep and swelling material models are introduced. Then, a theoretical solution based on a simplified bending plate model is briefly given. Finally, a stress update scheme for the adopted material models is presented and furthermore embedded into ABAQUS user interface UMAT to conduct finite element analysis. Both solutions are compared and discussed. Results: FEM successfully predicts the duct dilation and its solution agrees well with theoretical one in small deformation. Conclusions: The proposed stress update scheme is effective, The accuracy of the theory solution declines when dilation becomes larger. The maximum stress occurs at the duct corner point, and the location has stress relaxation effect. (authors)

Using Modeling and Simulation to Complement Testing for Increased Understanding of Weapon Subassembly Response.

Energy Technology Data Exchange (ETDEWEB)

Wong, Michael K. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Davidson, Megan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-09-01

As part of Sandia’s nuclear deterrence mission, the B61-12 Life Extension Program (LEP) aims to modernize the aging weapon system. Modernization requires requalification and Sandia is using high performance computing to perform advanced computational simulations to better understand, evaluate, and verify weapon system performance in conjunction with limited physical testing. The Nose Bomb Subassembly (NBSA) of the B61-12 is responsible for producing a fuzing signal upon ground impact. The fuzing signal is dependent upon electromechanical impact sensors producing valid electrical fuzing signals at impact. Computer generated models were used to assess the timing between the impact sensor’s response to the deceleration of impact and damage to major components and system subassemblies. The modeling and simulation team worked alongside the physical test team to design a large-scale reverse ballistic test to not only assess system performance, but to also validate their computational models. The reverse ballistic test conducted at Sandia’s sled test facility sent a rocket sled with a representative target into a stationary B61-12 (NBSA) to characterize the nose crush and functional response of NBSA components. Data obtained from data recorders and high-speed photometrics were integrated with previously generated computer models in order to refine and validate the model’s ability to reliably simulate real-world effects. Large-scale tests are impractical to conduct for every single impact scenario. By creating reliable computer models, we can perform simulations that identify trends and produce estimates of outcomes over the entire range of required impact conditions. Sandia’s HPCs enable geometric resolution that was unachievable before, allowing for more fidelity and detail, and creating simulations that can provide insight to support evaluation of requirements and performance margins. As computing resources continue to improve, researchers at Sandia are hoping

EBR-II: search for the lost subassembly

International Nuclear Information System (INIS)

King, R.W.; Buschman, H.W.; Poloncsik, J.; Remsburg, J.S.; Sine, H.W.

1983-01-01

Experimental Breeder Reactor II (EBR-II) has been operating for nearly 20 years as part of the foundation of the US Department of Energy's LMFBR development program. During that time, the EBR-II fuel-handling system has performed extremely well, especially considering the conditions under which much of the system operates and the reliability required to maintain the high plant factor routinely demonstrated by EBR-II. Since EBR-II is a pool-type reactor, much of the fuel handling is done remotely within the sodium-filled primary tank at 371 0 C. Activities involved in locating a misplaced fuel subassembly in the primary tank are described

Integrated circuit manufacture and tuning of subassemblies of a statistical analyzer of voltage oscillations (AKON). Izgotovleniye na integral'nykh skhemakh i nastroyka uzlov statisticheskogo analizatora kolebaniy napryazheniya (AKON)

Energy Technology Data Exchange (ETDEWEB)

Yermakov, V.F.; Oleynik, V.I.; Sambarov, Yu.M.

1982-01-01

The basic circuits and instructions for tuning subassemblies of a statistical analyzer of voltage oscillation are described. The device is intended for monitoring quality of voltage in electric networks in accordance with GOST13109-67. The component base of the device includes integrated circuits of the series 140, 155, 218 and 228.

Studies to single subassembly flow monitoring with a complete 7 element array under sodium

International Nuclear Information System (INIS)

Hess, B.; Ruppert, E.; Stehle, H.; Vinzens, K.

1975-01-01

As part of the SNR-300 R and D programme a complete clamped array, consisting of 4 full size fuel elements and 3 blanket elements was tested for more than 4000 hours at 600 deg C in the AKB sodium loop at Interatom, Bensberg. The test was split into two phases and the total cluster was prestrained in the second phase to simulate 15 mm subassembly displacement at the level of the upper bearing pads. Although this test was mainly considered as an endurance test to demonstrate the integrity of prestrained core elements, efforts were made to study the feasibility of single subassembly flow monitoring with this full size model of a core section. The results of these investigations are presented and discussed in this paper

Materials requirements for liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Bennett, J.W.; Horton, K.E.

1978-01-01

Materials requirements for Liquid Metal Fast Breeder Reactors (LMFBRs) are quite varied with requisite applications ranging from ex-reactor components such as piping, pumps, steam generators and heat exchangers to in-reactor components such as heavy section reactor vessels, core structurals, fuel pin cladding and subassembly flow ducts. Requirements for ex-reactor component materials include: good high temperature tensile, creep and fatigue properties; compatibility with high temperature flowing sodium; resistance to wear, stress corrosion cracking, and crack propagation; and good weldability. Requirements for in-reactor components include most of those cited above for ex-reactor components as supplemented by the following: resistance to radiation embrittlement, swelling and radiation enhanced creep; good neutronics; compatibility with fuel and fission product materials; and resistance to mass transfer via flowing sodium. Extensive programs are currently in place in a number of national laboratories and industrial contractors to address the materials requirements for LMFBRs. These programs are focused on meeting the near term requirements of early LMFBRs such as the Fast Flux Test Facility and the Clinch River Breeder Reactor as well as the longer term requirements of larger near-commercial and fully-commercial reactors

Pump and Flow Control Subassembly of Thermal Control Subsystem for Photovoltaic Power Module

Science.gov (United States)

Motil, Brian; Santen, Mark A.

1993-01-01

The pump and flow control subassembly (PFCS) is an orbital replacement unit (ORU) on the Space Station Freedom photovoltaic power module (PVM). The PFCS pumps liquid ammonia at a constant rate of approximately 1170 kg/hr while providing temperature control by flow regulation between the radiator and the bypass loop. Also, housed within the ORU is an accumulator to compensate for fluid volumetric changes as well as the electronics and firmware for monitoring and control of the photovoltaic thermal control system (PVTCS). Major electronic functions include signal conditioning, data interfacing and motor control. This paper will provide a description of each major component within the PFCS along with performance test data. In addition, this paper will discuss the flow control algorithm and describe how the nickel hydrogen batteries and associated power electronics will be thermally controlled through regulation of coolant flow to the radiator.

Reconstitutable nuclear reactor fuel assembly with unitary removable top nozzle subassembly

International Nuclear Information System (INIS)

Shallenberger, J.M.

1987-01-01

A reconstitutable fuel assembly is described having at least one control rod guide thimble and a top nozzle, the guide thimble including an upper extension, the top nozzle including at least one hold-down spring, an upper hold-down plate and a lower adapter plate, an improved attaching structure removably mounting the top nozzle as a unitary subassembly on the guide thimble. The attaching structure comprises: (a) a coupling member interfitting the lower adapter plate, the upper hold-down plate and the hold-down spring disposed between the plates so as to capture and retain the plates and spring together as a unitary subassembly in which the upper plate is slidably moveable along the coupling member relative to the lower plate with the spring biasing the upper plate away from the lower plate. The coupling member has spaced apart upper and lower portions with a central passageway extending for slidably receiving the upper extension of the guide thimble in a nonattached relationship in which the coupling member is slidably movable relative to the guide thimble extension for respectively inserting and removing the coupling member on and from the guide thimble extension

Fuel cell subassemblies incorporating subgasketed thrifted membranes

Science.gov (United States)

Iverson, Eric J.; Pierpont, Daniel M.; Yandrasits, Michael A.; Hamrock, Steven J.; Obradovich, Stephan J.; Peterson, Donald G.

2016-03-01

A fuel cell roll good subassembly is described that includes a plurality of individual electrolyte membranes. One or more first subgaskets are attached to the individual electrolyte membranes. Each of the first subgaskets has at least one aperture and the first subgaskets are arranged so the center regions of the individual electrolyte membranes are exposed through the apertures of the first subgaskets. A second subgasket comprises a web having a plurality of apertures. The second subgasket web is attached to the one or more first subgaskets so the center regions of the individual electrolyte membranes are exposed through the apertures of the second subgasket web. The second subgasket web may have little or no adhesive on the subgasket surface facing the electrolyte membrane.

Development of automatic gamma and neutron monitoring system for PFBR fuel subassemblies at IFSB

International Nuclear Information System (INIS)

Krishnakumar, D.N.; Dhanasekaran, A.; Ajoy, K.C.; Jose, M.T.; Baskaran, R.; Sureshkumar, K.V.

2018-01-01

Health physics surveillance during PFBR fuel pin assembling operation at Interim Fuel Storage Building (IFSB) mandates scanning of the fuel assembly using Telector and Rem counter to find out the maximum gamma and neutron dose rates respectively. Throughout the process health physicist involved in the operation must hold the survey meter at a constant distance from the subassembly and simultaneously should make a note of dose rate values displayed. This practice might lead to the occupational exposures and also might induce human errors during measurements. To make this process more simple, faultless and effortless, an automatic Gamma Neutron Monitoring System (AGNMS) is designed and developed at RSD to measure, store and visualize instantaneous gamma and neutron dose rates of PFBR fuel subassembly. Development of the system, calibration and deployment of the system at IFSB and preliminary results obtained using the system is depicted in this paper

Seismic analysis and design of spent subassembly storage bay (SSSB) pool

International Nuclear Information System (INIS)

Abdul Gani, H.I.; Ramanjaneyulu, K.V.S.; Pillai, C.S.; Chetal, S.C.

2003-01-01

Fuel bundles, after their specified stay in reactor core, are replaced by fresh fuel for sustaining power generation at rated levels. The irradiated fuel subassembly, removed fresh from core, known as spent fuel sub assembly, is radioactive and decay heat generating. It needs to be cooled before it becomes amenable for handling, either for reprocessing or for immobilisation. For this purpose, it is immersed in a pool of water, retained in a concrete structure referred as Spent Subassembly Storage Bay (SSSB) pool. The height of water column above fuel bundles is arrived from shielding considerations. SSSB pool is one of the nuclear safety related structures and warrants rigorous analysis and design. The SSSB pool, in case of PFBR 500 MW(e) is located in fuel building. It is a stainless steel lined. water retaining rectangular R.C.C. open tank of size 7.5 X 29.0 m, with a height of 11.0 m. This structure is analysed for two levels of site specific earthquakes taking in to account liquid structure interactions as per ASCE-4, 1998. The design of walls and bottom slab is carried out satisfying the AERB code for nuclear safety related structures. Analysis and design of SSSB pool of PFBR is presented in the following paper. (author)

Field test and evaluation of the passive neutron coincidence collar for prototype fast reactor fuel subassemblies

International Nuclear Information System (INIS)

Menlove, H.O.; Keddar, A.

1982-08-01

The passive neutron Coincidence Collar, which was developed for the verification of plutonium content in fast reactor fuel subassemblies, has been field tested using Prototype Fast Reactor fuel. For passive applications, the system measures the 240 Pu-effective mass from the spontaneous fission rate, and in addition, a self-interrogation technique is used to determine the fissile content in the subassembly. Both the passive and active modes were evaluated at the Windscale Works in the United Kingdom. The results of the tests gave a standard deviation 0.75% for the passive count and 3 to 7% for the active measurement for a 1000-s counting time. The unit will be used in the future for the verification of plutonium in fresh fuel assemblies

Thermal experiments with LMFBR subassembly models in sodium flow

International Nuclear Information System (INIS)

Moeller, R.; Tschoeke, H.

1982-01-01

Within the framework of the Fast Breeder Project research work has been undertaken at the Karlsruhe Nuclear Research Center on the thermal and fluid dynamics of nominal and distorted core subassemblies. In 19-rod bundle models (P/D=1.30, W/R=1.38) three-dimensional temperature distributions were measured in the cladding tubes exposed to sodium flow. Results of measurements of the azimuthal temperature profiles of rotated rods in the duct wall zone are indicated for different operating conditions 80 2 , evenly distributed load and oblique load; different axial positions of the spacer grids; and different positions of one bowed rod

Manufacture of core sub-assemblies and fertile fuel assemblies for Indian fast breeder programme

International Nuclear Information System (INIS)

Jayaraj, R.N.

2009-01-01

sintered ThO 2 . Different varieties of stainless steel are employed for the manufacture of intricate components required for Fast Breeder sub-assemblies which are not easily machinable. Large number of precision machined components are fabricated through specialized machining, forming and welding techniques and finally assembled with the help of special jigs and fixtures. Several fabrication techniques were developed like Clad Tube Crimping, Button Forming of Hexagonal Tube, Bead Forming of Spacer Wire, Welding of Clad Tubes to End Plugs and Hexagonal Tube to Foot and Handling Head. Specialized Joining Techniques like Pulsed Current GTAW are employed for the fabrication of thin- walled Fuel Elements. Developmental works are also undertaken for standardizing manufacturing techniques for Oxide Dispersion Strengthened (ODS) alloys for clad tubes of Fast Breeder Reactors, which will have an edge over conventional materials with respect to excellent resistance to void swelling and irradiation embrittlement and also capable of operating under severe conditions for extended periods. The paper highlights various developmental activities carried out for the manufacture of core sub-assemblies for the Fast Breeder Test Reactor (FBTR) and the forthcoming Prototype Fast Breeder Reactor (PFBR). (author)

EBRPOCO - a program to calculate detailed contributions of power reactivity components of EBR-II

International Nuclear Information System (INIS)

Meneghetti, D.; Kucera, D.A.

1981-01-01

The EBRPOCO program has been developed to facilitate the calculations of the power coefficients of reactivity of EBR-II loadings. The program enables contributions of various components of the power coefficient to be delineated axially for every subassembly. The program computes the reactivity contributions of the power coefficients resulting from: density reduction of sodium coolant due to temperature; displacement of sodium coolant by thermal expansions of cladding, structural rods, subassembly cans, and lower and upper axial reflectors; density reductions of these steel components due to temperature; displacement of bond-sodium (if present) in gaps by differential thermal expansions of fuel and cladding; density reduction of bond-sodium (if present) in gaps due to temperature; free axial expansion of fuel if unrestricted by cladding or restricted axial expansion of fuel determined by axial expansion of cladding. Isotopic spatial contributions to the Doppler component my also be obtained. (orig.) [de

French studies on local blockages in LMFBR fuel subassemblies

International Nuclear Information System (INIS)

Girard, C.; Jolas, P.; Seiler, J.M.

1979-08-01

This paper reviews experimental and theoretical studies done in FRANCE on the problem of partial subassembly blockages. The priorities are defined and the development of the French program in the European context is presented. Results of the out of pile experiments performed at CEA and EDF in single and two phases flow are given. A description of the main codes used to interpret these experiments is then shortly reviewed. It is found that the thermal behavior in single phase may be calculated with good precision, and that a simple semi-empirical formula can predict with good accuracy the number of channels blocked that lead to sodium boiling

Methods for nuclear material control used in the basic production of a typical radiochemical plant

International Nuclear Information System (INIS)

Kositsyn, V.F.; Mukhortov, N.F.; Korovin, Yu.I.; Rudenko, V.S.; Petrov, A.M.

1999-01-01

Techniques for destructive and non-destructive assay of the component and isotopic composition of nuclear materials are described, namely gravimetric, titrimetric, coulometric, mass spectrometry, as well as those based on registration of neutron and γ radiations. Their metrologic characteristics are described. The techniques described are suggested to be used for nuclear material (NM) control and accounting purposes at the model radiochemical plant for processing irradiated fuel subassemblies from power reactors. The measurement control program is also described. This program is intended for the measurement quality assurance in the framework of NM control and accountancy system [ru

Flow of ideal fluid through a central region of a nuclear reactor wire-spaced fuel subassembly

International Nuclear Information System (INIS)

Schmid, J.

1991-04-01

The results are given of calculations of the flow of an ideal fluid through the central region of a nuclear reactor wire-spaced fuel subassembly. The computer code used is briefly described. (author). 10 figs., 4 refs

Subassembly faults diagnostic of an LMFBR type reactor by the measurement of temperature noise

International Nuclear Information System (INIS)

Kokorev, B.V.; Palkin, I.I.; Turchin, N.M.; Pallagi, D.; Horanyi, S.

1979-09-01

The subassembly faults detection possibility by temperature noise analysis of an LMFBR is described. The paper contains the results of diagnostical examinations obtained on electrically heated NaK test rigs. On the basis of these results the measurement of temperature noise RMS value seems to be a practicable method to detect local blockages in an early phase. (author)

Improvement of computer programs 'BAMBOO' and 'ASFRE-IV' for coupling analysis of deformation and thermal-hydraulics in a high burn-up fuel subassembly of fast reactor

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ohshima, Hiroyuki; Imai, Yasutomo

2003-04-01

A simulation system of a deformed fuel subassembly is being developed for the structure integrity of high burn-up wire-spacer-type fuel subassemblies of sodium-cooled fast breeder reactors. This report describes a computer program improvement work for coupling analyses of deformation and thermal-hydraulics in a fuel subassembly as part of the simulation system development. In this work, a function of data conversion as an interface between a bundle deformation analysis program BAMBOO and a thermal hydraulic analysis program ASFRE-IV was incorporated to each program. BAMBOO was improved to accept the coolant temperature data from ASFRE-IV and to offer bundle deformation data to ASFRE-IV. ASFRE-IV was also improved to offer the coolant temperature data to BAMBOO and to obtain the bundle deformation data from BAMBOO. Improved BAMBOO and ASFRE-IV were applied to an analysis of 169-pin bundle for the program verification. It was confirmed that the coupling analysis gave the physically reasonable results on both deformation and thermal hydraulic behaviors in the fuel subassembly. (author)

Development of a FBR fuel bundle-duct interaction analysis code-BAMBOO. Analysis model and verification by Phenix high burn-up fuel subassemblies

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ito, Masahiro; Ukai, Shigeharu

2005-01-01

The bundle-duct interaction analysis code ''BAMBOO'' has been developed for the purpose of predicting deformation of a wire-wrapped fuel pin bundle of a fast breeder reactor (FBR). The BAMBOO code calculates helical bowing and oval-distortion of all the fuel pins in a fuel subassembly. We developed deformation models in order to precisely analyze the irradiation induced deformation by the code: a model to analyze fuel pin self-bowing induced by circumferential gradient of void swelling as well as thermal expansion, and a model to analyze dispersion of the orderly arrangement of a fuel pin bundle. We made deformation analyses of high burn-up fuel subassemblies in Phenix reactor and compared the calculated results with the post irradiation examination data of these subassemblies for the verification of these models. From the comparison we confirmed that the calculated values of the oval-distortion and bowing reasonably agreed with the PIE results if these models were used in the analysis of the code. (author)

Preliminary validation of the MATRA-LMR-FB code for the flow blockage in a subassembly

International Nuclear Information System (INIS)

Jeong, H. Y.; Ha, K. S.; Kwon, Y. M.; Chang, W. P.; Lee, Y. B.; Heo, S.

2005-01-01

To analyze the flow blockage in a subassembly of a Liquid Metal-cooled Reactor (LMR), the MATRA-LMR-FB code has been developed and validated for the existing experimental data. Compared to the MATRA-LMR code, which had been successfully applied for the core thermal-hydraulic design of KALIMER, the MATRA-LMR-FB code includes some advanced modeling features. Firstly, the Distributed Resistance Model (DRM), which enables a very accurate description of the effects of wire-wrap and blockage in a flow path, is developed for the MATRA-LMR-FB code. Secondly, the hybrid difference method is used to minimize the numerical diffusion especially at the low flow region such as recirculating wakes after blockage. In addition, the code is equipped with various turbulent mixing models to describe the active mixing due to the turbulent motions as accurate as possible. For the validation of the MATRA-LMR-FB code the ORNL THORS test and KOS 169-pin test are analyzed. Based on the analysis results for the temperature data, the accuracy of the code is evaluated quantitatively. The MATRA-LMR-FB code predicts very accurately the exit temperatures measured in the subassembly with wire-wrap. However, the predicted temperatures for the experiment with spacer grid show some deviations from the measured. To enhance the accuracy of the MATRA-LMR-FB for the flow path with grid spacers, it is suggested to improve the models for pressure loss due to spacer grid and the modeling method for blockage itself. The developed MATRA-LMR-FB code is evaluated to be applied to the flow blockage analysis of KALIMER-600 which adopts the wire-wrapped subassemblies

Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

International Nuclear Information System (INIS)

2005-01-01

This Working Material includes the papers presented at the International Meeting 'Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two-phase', which was held 5-9 July 2004 at the State Scientific Center of Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky, in Obninsk near Moscow. The objectives of the meeting were to discuss new results obtained in the field of liquid metal coolant and to recommend the lines of further general physics and applied investigations, with the purpose of validating existing and codes under development for liquid metal cooled advanced and new generation nuclear reactors. Most of the contributions present results of experimental and numerical investigations into velocity, temperature and heat transfer in fuel subassemblies of fast reactors cooled by sodium or lead. In the frame of the meeting a benchmark problem devoted to heat transfer in the model subassembly of the fast reactor BREST-OD-300 was proposed. Experts from 5 countries (Japan, Netherlands, Spain, Republic of Korea, and Russia) took part in this benchmark exercise. The results of the benchmark calculations are summarized in the Working Material. The results of hydrodynamic studies of pressure head chambers and collector systems of liquid metal cooled reactors are presented in a number of papers. Also attention was given to the generalization of experimental data on hydraulic losses in the pipelines in case of mutual influence of local pressure drops, and to the modeling of natural convection in the fuel subassemblies and circuits with liquid metal cooling. Special emphasis at the meeting was placed on thermal hydraulics issues related to the development and design of target systems, such as heat removal in the target unit of the cascade subcritical reactor cooled by liquid salt; the target complex MK-1 for accelerator driven systems cooled by eutectic lead-bismuth alloy; and the test

Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-07-01

This Working Material includes the papers presented at the International Meeting 'Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two-phase', which was held 5-9 July 2004 at the State Scientific Center of Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky, in Obninsk near Moscow. The objectives of the meeting were to discuss new results obtained in the field of liquid metal coolant and to recommend the lines of further general physics and applied investigations, with the purpose of validating existing and codes under development for liquid metal cooled advanced and new generation nuclear reactors. Most of the contributions present results of experimental and numerical investigations into velocity, temperature and heat transfer in fuel subassemblies of fast reactors cooled by sodium or lead. In the frame of the meeting a benchmark problem devoted to heat transfer in the model subassembly of the fast reactor BREST-OD-300 was proposed. Experts from 5 countries (Japan, Netherlands, Spain, Republic of Korea, and Russia) took part in this benchmark exercise. The results of the benchmark calculations are summarized in the Working Material. The results of hydrodynamic studies of pressure head chambers and collector systems of liquid metal cooled reactors are presented in a number of papers. Also attention was given to the generalization of experimental data on hydraulic losses in the pipelines in case of mutual influence of local pressure drops, and to the modeling of natural convection in the fuel subassemblies and circuits with liquid metal cooling. Special emphasis at the meeting was placed on thermal hydraulics issues related to the development and design of target systems, such as heat removal in the target unit of the cascade subcritical reactor cooled by liquid salt; the target complex MK-1 for accelerator driven systems cooled by eutectic lead-bismuth alloy; and the test

Fuels and materials research under the high neutron fluence using a fast reactor Joyo and post-irradiation examination facilities

International Nuclear Information System (INIS)

Soga, Tomonori; Ito, Chikara; Aoyama, Takafumi; Suzuki, Soju

2009-01-01

The experimental fast reactor Joyo at Oarai Research and Development Center (ORDC) of Japan Atomic Energy Agency (JAEA) is Japan's sodium-cooled fast reactor (FR). In 2003, this reactor's upgrade to the 140MWt MK-III core was completed to increase the irradiation testing capability. The MK-III core provides the fast neutron flux of 4.0x10 15 n/cm 2 s as an irradiation test bed for improving the fuels and material of FR in Japan. Three post-irradiation examination (PIE) facilities named FMF, MMF and AGF related to Joyo are in ORDC. Irradiated subassemblies and core components are carried into the FMF (Fuel Monitoring Facility) and conducted nondestructive examinations. Each subassembly is disassembled to conduct some destructive examinations and to prepare the fuel and material samples for further detailed examinations. Fuel samples are sent to the AGF (Alpha-Gamma Facility), and material samples are sent to the MMF (Materials Monitoring Facility). These overall and elaborate data provided by PIE contribute to investigate the irradiation effect and behavior of fuels and materials. This facility complex is indispensable to promote the R and D of FR in Japan. And, the function and technology of irradiation test and PIE enable to contribute to the R and D of innovative fission or fusion reactor material which will be required to use under the high neutron exposure. (author)

Device and method for unfastening and lifting a top nozzle subassembly from a reconstitutable fuel assembly

International Nuclear Information System (INIS)

Wilson, J.F.

1987-01-01

This patent describes a reconstitutable fuel assembly including at least one guide thimble having an upper end portion and a top nozzle subassembly having a lower adapter plate with at least one opening, and an upper hold-down plate with at least one passageway positioned above and aligned with the lower adapter plate opening. At least one hold-down spring is disposed and extends between the upper and lower plates and at least one elongated tubular hollow sleeve is disposed and extends between the upper and lower plates. The upper end portion of the guide thimble extends upwardly through the opening in the lower adapter plate and has a threaded terminal end disposed above the adapter plate. The threaded terminal end of the guide thimble and an upper end extend upwardly through the passageway of the upper hold-down plate. A device is described for unfastening and lifting the top nozzle subassembly from the guide thimble of the fuel assembly, comprising: (a) at least one hollow gripper tube, the tube having an open lower end; (b) means mounting the gripper tube for vertical alignment with and insertion of its lower end portion into the elongated sleeve of the top nozzle subassembly to a position therein located above and adjacent to the threaded lower end of the sleeve; (c) force-generating means disposed within the gripper tube for rotatable movement and concurrent axial movement upwardly and downwardly within the tube and also disposed at the open lower end of the gripper tube for extension into and from the gripper tube open lower end upon axial movement upwardly and downwardly within the gripper tube

Laser Beam Melting of Multi-Material Components

Science.gov (United States)

Laumer, Tobias; Karg, Michael; Schmidt, Michael

First results regarding the realisation of multi-material components manufactured by Laser Beam Melting of polymers and metals are published. For realising composite structures from polymer powders by additive manufacturing, at first relevant material properties regarding compatibility have to be analysed. The paper shows the main requirements for compatibility between different materials and offers first results in form of a compatibility matrix of possible combinations for composite structures. For achieving gradient properties of additively manufactured metal parts by using composite materials the composition of alloying components in the powder and adapted process strategies are varied. As an alternative to atomizing pre-alloyed materials, mixtures of different powders are investigated.

Passive RF component technology materials, techniques, and applications

CERN Document Server

Wang, Guoan

2012-01-01

Focusing on novel materials and techniques, this pioneering volume provides you with a solid understanding of the design and fabrication of smart RF passive components. You find comprehensive details on LCP, metal materials, ferrite materials, nano materials, high aspect ratio enabled materials, green materials for RFID, and silicon micromachining techniques. Moreover, this practical book offers expert guidance on how to apply these materials and techniques to design a wide range of cutting-edge RF passive components, from MEMS switch based tunable passives and 3D passives, to metamaterial-bas

Irradiation experience with KNK II Fast Breeder Fuel Subassemblies

International Nuclear Information System (INIS)

Hess, B.

1993-02-01

During the operation of the second core of KNK II fuel pin failures occurred, which were caused by local cladding weakening due to mechanical interaction between fuel pins and pin spacers. The present report gives a summarizing presentation of the consequences of these interactions, of the experimental and theoretical investigations to clarify the reason for the interactions and of measures to reduce their consequences in the extended residence time of the second core of KNK II. This type of interaction is caused by thermo-elastic instabilities of the fuel pin bundle, and its strength depends sensitively on the geometry of the pin bundle and the pin power. Finally, measures are described, which were taken for the fuel subassemblies of the third core of KNK II to avoid the wear causing instabilities [de

Comparisons of PRD [power-reactivity-decrements] components for various EBR-II configurations

International Nuclear Information System (INIS)

Meneghetti, D.; Kucera, D.A.

1986-01-01

Comparison of detailed calculations of contributions by region and component of the power-reactivity-decrements (PRD) for four differing loading configurations of the Experimental Breeder Reactor-II (EBR-II) are given. The linear components and Doppler components are calculated. The non-linear (primarily subassembly bowing) components are deduced by differences relative to measured total PRD values. Variations in linear components range from about 10% to as much as about 100% depending upon the component. The deduced non-linear components differ both in magnitude and sign as functions of reactor power. Effects of differing assumptions of the nature of the fuel-to-clad interactions upon the PRD components are also calculated

Probabilistic distributions of pin gaps within a wire-spaced fuel subassembly and sensitivities of the related uncertainties to pin gap

International Nuclear Information System (INIS)

Sakai, K.; Hishida, H.

1978-01-01

Probabilistic fuel pin gap distributions within a wire-spaced fuel subassembly and sensitivities of the related uncertainties to fuel pin gaps are discussed. The analyses consist mainly of expressing a local fuel pin gap in terms of sensitivity functions of the related uncertainties and calculating the corresponding probabilistic distribution through taking all the possible combinations of the distribution of uncertainties. The results of illustrative calculations show that with the reliability level of 0.9987, the maximum deviation of the pin gap at the cladding hot spot of a center fuel subassembly is 8.05% from its nominal value and the corresponding probabilistic pin gap distribution is shifted to the narrower side due to the external confinement of a pin bundle with a wrapper tube. (Auth.)

Durability of building materials and components

CERN Document Server

Delgado, JMPQ

2013-01-01

Durability of Building Materials and Components provides a collection of recent research works to contribute to the systematization and dissemination of knowledge related to the long-term performance and durability of construction and, simultaneously, to show the most recent advances in this domain. It includes a set of new developments in the field of durability, service life prediction methodologies, the durability approach for historical and old buildings, asset and maintenance management and on the durability of materials, systems and components. The book is divided in several chapters that intend to be a resume of the current state of knowledge for benefit of professional colleagues.

Applications of ultrasonic phased array technique during fabrication of nuclear tubing and other components for the Indian nuclear power program

International Nuclear Information System (INIS)

Kapoor, K.

2015-01-01

Ultrasonic phased array technique has been applied in fabrication of nuclear fuel and structural at NFC. The integrity of the nuclear fuel and structural components is most crucial as they are exposed to severe environment during operation leading to rapid degradation of its properties during its lifecycle. Nuclear Fuel Complex has mandate for the fabrication of the nuclear fuel and core structurals for Indian PHWRs/BWR, sub-assemblies for the PFBR and steam generator tubing for PFBR and PHWRs which are the most critical materials for the Indian Nuclear Power program. NDE during fabrication of these materials is thus most crucial as it provides the confidence to the designer for safe operation during its lifetime. Many of these techniques have to be developed in-house to meet unique requirements of high sensitivity, resolution and shape of the components. Some of the advancements in the NDE during the fabrication include use of ultrasonic phased array which is detailed in this paper

Quasi-steady state boiling downstream of a central blockage in a 19-rod simulated LMFBR subassembly (FFM bundle 3B)

International Nuclear Information System (INIS)

Hanus, N.; Fontana, M.H.; Gnadt, P.A.; MacPherson, R.E.; Smith, C.M.; Wantland, J.L.

1976-01-01

Results of sodium boiling tests in a centrally blocked 19-rod simulated LMFBR subassembly are discussed. The tests were part of the experimental series conducted with bundle 3B in the Fuel Failure Mockup (FFM) at ORNL

Reliability & availability of wind turbine electrical & electronic components

NARCIS (Netherlands)

Tavner, P.; Faulstich, S.; Hahn, B.; Bussel, van G.J.W.

2010-01-01

Recent analysis of European onshore wind turbine reliability data has shown that whilst wind turbine mechanical subassemblies tend to have relatively low failure rates but long downtimes, electrical and electronic subassemblies have relatively high failure rates and short downtimes. For onshore wind

Computation of turbulent flow and heat transfer in subassemblies

International Nuclear Information System (INIS)

Slagter, W.

1979-01-01

This research is carried out in order to provide information on the thermohydraulic behaviour of fast reactor subassemblies. The research work involves the development of versatile computation methods and the evaluation of combined theoretical and experimental work on fluid flow and heat transfer in fuel rod bundles. The computation method described here rests on the application of the distributed parameter approach. The conditions considered cover steady, turbulent flow and heat transfer of incompressible fluids in bundles of bare rods. Throughout 1978 main efforts were given to the development of the VITESSE program and to the validation of the hydrodynamic part of the code. In its present version the VITESSE program is applicable to predict the fully developed turbulent flow and heat transfer in the subchannels of a bundle with bare rods. In this paper the main features of the code are described as well as the present status of development

Advanced Electrical Materials and Components Development: An Update

Science.gov (United States)

Schwarze, Gene E.

2005-01-01

The primary means to develop advanced electrical components is to develop new and improved materials for magnetic components (transformers, inductors, etc.), capacitors, and semiconductor switches and diodes. This paper will give an update of the Advanced Power Electronics and Components Technology being developed by the NASA Glenn Research Center for use in future Power Management and Distribution subsystems used in space power systems for spacecraft and lunar and planetary surface power. The initial description and status of this technology program was presented two years ago at the First International Energy Conversion Engineering Conference held at Portsmouth, Virginia, August 2003. The present paper will give a brief background of the previous work reported and a summary of research performed the past several years on soft magnetic materials characterization, dielectric materials and capacitor developments, high quality silicon carbide atomically smooth substrates, and SiC static and dynamic device characterization under elevated temperature conditions. The rationale for and the benefits of developing advanced electrical materials and components for the PMAD subsystem and also for the total power system will also be briefly discussed.

76 FR 45845 - Notice of Issuance of Final Determination Concerning a Certain Patient Transport Chair

Science.gov (United States)

2011-08-01

... approximately 481 components. All of the components are of U.S., Chinese, Canadian, or French origin. The... (which includes a French-origin handle circuit board, a control box, a key switch subassembly, and a... battery cable subassemblies, a handle cable subassembly, an emergency stop switch subassembly, a horn...

EXPEL - a computing module for molten fuel/coolant interactions in fast reactor sub-assemblies

International Nuclear Information System (INIS)

Fishlock, T.P.

1975-10-01

This report describes a module for computing the effects of a molten fuel/coolant interaction in a fast reactor subassembly. The module is to be incorporated into the FRAX code which calculates the consequences of hypothetical whole core accidents. Details of the interaction are unknown and in consequence the model contains a large number of parameters which must be set by assumption. By variation of these parameters the interaction may be made mild or explosive. Results of a parametric survey are included. (author)

The EBR-II materials-surveillance program. 4: Results of SURV-4 and SURV-6

International Nuclear Information System (INIS)

Ruther, W.E.; Hayner, G.O.; Carlson, B.G.; Ebersole, E.R.; Allen, T.R.

1998-01-01

In March of 1965, a set of surveillance (SURV) samples was placed in the EBR-II reactor to determine the effect of irradiation, thermal aging, and sodium corrosion on reactor materials. Eight subassemblies were placed into row 12 positions of EBR-II to determine the effect of irradiation at 370 C. Two subassemblies were placed into the primary sodium basket to determine the effect of thermal aging at 370 C. For both the irradiated and thermally aged samples, one half of all samples were exposed to primary system sodium while one half were sealed in capsules with a helium atmosphere. Fifteen different structural materials were tested in the SURV program. In addition to the fifteen types of metal samples, graphite blocks were irradiated in the SURV subassemblies to determine the effect of irradiation on the graphite neutron shield. In this report, the properties of these materials irradiated at 370 C to a total fluence of 2.2 x 10 22 n/cm 2 (over 2,994 days) are compared with those of similar specimens thermally aged at 370 C for 2,994 days in the storage basket of the reactor. The properties analyzed were weight, density, microstructure, hardness, tensile and yield strength, impact strength, and creep

Simulation software of 3-D two-neutron energy groups for ship reactor with hexagonal fuel subassembly

International Nuclear Information System (INIS)

Zhang Fan; Cai Zhangsheng; Yu Lei; Gui Xuewen

2005-01-01

Core simulation software for 3-D two-neutron energy groups is developed. This software is used to simulate the ship reactor with hexagonal fuel subassembly after 10, 150 and 200 burnup days, considering the hydraulic and thermal feedback. It accurately simulates the characteristics of the fast and thermal neutrons and the detailed power distribution in a reactor under normal and abnormal operation condition. (authors)

Large scale FCI experiments in subassembly geometry. Test facility and model experiments

International Nuclear Information System (INIS)

Beutel, H.; Gast, K.

A program is outlined for the study of fuel/coolant interaction under SNR conditions. The program consists of a) under water explosion experiments with full size models of the SNR-core, in which the fuel/coolant system is simulated by a pyrotechnic mixture. b) large scale fuel/coolant interaction experiments with up to 5kg of molten UO 2 interacting with liquid sodium at 300 deg C to 600 deg C in a highly instrumented test facility simulating an SNR subassembly. The experimental results will be compared to theoretical models under development at Karlsruhe. Commencement of the experiments is expected for the beginning of 1975

Analyses of deformation and thermal-hydraulics within a wire-wrapped fuel subassembly in a liquid metal fast reactor by the coupled code system

Energy Technology Data Exchange (ETDEWEB)

Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp; Ohshima, Hiroyuki; Ito, Masahiro

2017-06-15

Highlights: • The coupled computational code system allowed for mechanical and thermal-hydraulic analyses in a fast reactor fuel subassembly. • In this system interactive calculations between flow area deformations and coolant temperature changes are repeated to their convergence state. • Effects on bundle-duct interaction on coolant temperature distributions were investigated by using the code system. - Abstract: The coupled numerical analysis of mechanical and thermal-hydraulic behaviors was performed for a wire-wrapped fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal-hydraulic analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that the radial distribution of coolant temperature in the subassembly tended to flatten as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such flattening of temperature distribution was slightly observed as a result of fuel pin bowings due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal-hydraulics was also investigated in this study.

Material and construction of primary components

International Nuclear Information System (INIS)

Kaser, A.; Wallner, F.

1978-01-01

The construction of SNR's requires specific properties of the materials, i.e. high strength at temperatures of 600 0 C, adequate creep rupture strength, low long-time embrittlement. Aspects are given for optimalization of the mentioned properties with regard to safe manufacture especially good weldability. The austenitic material X6CrNil811 similar the type AISI 304 SS finally was chosen. Besides the fundamental analysis of the material properties it will be reported about the experiences gained during the manufacturing of the essential components. (author)

Mechanical Components from Highly Recoverable, Low Apparent Modulus Materials

Science.gov (United States)

Padula, Santo, II (Inventor); Noebe, Ronald D. (Inventor); Stanford, Malcolm K. (Inventor); DellaCorte, Christopher (Inventor)

2015-01-01

A material for use as a mechanical component is formed of a superelastic intermetallic material having a low apparent modulus and a high hardness. The superelastic intermetallic material is conditioned to be dimensionally stable, devoid of any shape memory effect and have a stable superelastic response without irrecoverable deformation while exhibiting strains of at least 3%. The method of conditioning the superelastic intermetallic material is described. Another embodiment relates to lightweight materials known as ordered intermetallics that perform well in sliding wear applications using conventional liquid lubricants and are therefore suitable for resilient, high performance mechanical components such as gears and bearings.

Materials and Components Technology Division research summary, 1992

International Nuclear Information System (INIS)

1992-11-01

The Materials and Components Technology Division (MCT) provides a research and development capability for the design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs related to nuclear energy support the development of the Integral Fast Reactor (IFR): life extension and accident analyses for light water reactors (LWRs); fuels development for research and test reactors; fusion reactor first-wall and blanket technology; and safe shipment of hazardous materials. MCT Conservation and Renewables programs include major efforts in high-temperature superconductivity, tribology, nondestructive evaluation (NDE), and thermal sciences. Fossil Energy Programs in MCT include materials development, NDE technology, and Instrumentation design. The division also has a complementary instrumentation effort in support of Arms Control Technology. Individual abstracts have been prepared for the database

The thermalhydraulics of a pin bundle with a helical wire wrap spacer. Modeling and qualification for a new sub-assembly concept

International Nuclear Information System (INIS)

Valentin, B.

2000-01-01

For the sub-assembly composed by an hexcan and a pin bundle with an helical wire wrap spacer, the calculation of the maximum clad temperatures, with the design code CADET, imposed to correctly evaluate the heat and mass transfers due to the helical wire. The models use theoretical and experimental arguments which are presented after a brief description of the hydraulic behavior of a such bundle. The design of a new sub-assembly concept, in the framework of high plutonium consumption in fast reactor projects needs to qualify tile models from RAPSODIE, PHENIX and SUPER-PHENIX programs. The qualification program, which could be used, is described. the approach is notably comparative for the hydraulic fields and the past experimental results will be useful. Another approach is briefly presented. It uses a multidimensional code (TRIO) which solves Navier-Stokes equations. The utility and the limits of a such method are described. (author)

Thermal-hydraulic numerical simulation of fuel sub-assembly for Sodium-cooled Fast Reactor

International Nuclear Information System (INIS)

Saxena, Aakanksha

2014-01-01

The thesis focuses on the numerical simulation of sodium flow in wire wrapped sub-assembly of Sodium-cooled Fast Reactor (SFR). First calculations were carried out by a time averaging approach called RANS (Reynolds- Averaged Navier-Stokes equations) using industrial code STAR-CCM+. This study gives a clear understanding of heat transfer between the fuel pin and sodium. The main variables of the macroscopic flow are in agreement with correlations used hitherto. However, to obtain a detailed description of temperature fluctuations around the spacer wire, more accurate approaches like LES (Large Eddy Simulation) and DNS (Direct Numerical Simulation) are clearly needed. For LES approach, the code TRIO U was used and for the DNS approach, a research code was used. These approaches require a considerable long calculation time which leads to the need of representative but simplified geometry. The DNS approach enables us to study the thermal hydraulics of sodium that has very low Prandtl number inducing a very different behavior of thermal field in comparison to the hydraulic field. The LES approach is used to study the local region of sub-assembly. This study shows that spacer wire generates the local hot spots (∼20 C) on the wake side of spacer wire with respect to the sodium flow at the region of contact with the fuel pin. Temperature fluctuations around the spacer wire are low (∼1 C-2 C). Under nominal operation, the spectral analysis shows the absence of any dominant peak for temperature oscillations at low frequency (2-10 Hz). The obtained spectra of temperature oscillations can be used as an input for further mechanical studies to determine its impact on the solid structures. (author) [fr

Assessment and selection of materials for ITER in-vessel components

Science.gov (United States)

Kalinin, G.; Barabash, V.; Cardella, A.; Dietz, J.; Ioki, K.; Matera, R.; Santoro, R. T.; Tivey, R.; ITER Home Teams

2000-12-01

During the international thermonuclear experimental reactor (ITER) engineering design activities (EDA) significant progress has been made in the selection of materials for the in-vessel components of the reactor. This progress is a result of the worldwide collaboration of material scientists and industries which focused their effort on the optimisation of material and component manufacturing and on the investigation of the most critical material properties. Austenitic stainless steels 316L(N)-IG and 316L, nickel-based alloys Inconel 718 and Inconel 625, Ti-6Al-4V alloy and two copper alloys, CuCrZr-IG and CuAl25-IG, have been proposed as reference structural materials, and ferritic steel 430, and austenitic steel 304B7 with the addition of boron have been selected for some specific parts of the ITER in-vessel components. Beryllium, tungsten and carbon fibre composites are considered as plasma facing armour materials. The data base on the properties of all these materials is critically assessed and briefly reviewed in this paper together with the justification of the material selection (e.g., effect of neutron irradiation on the mechanical properties of materials, effect of manufacturing cycle, etc.).

Development of multi-dimensional analysis method for porous blockage in fuel subassembly. Numerical simulation for 4 subchannel geometry water test

International Nuclear Information System (INIS)

Tanaka, Masa-aki; Kamide, Hideki

2001-02-01

This investigation deals with the porous blockage in a wire spacer type fuel subassembly in Fast Breeder Reactors (FBR's). Multi-dimensional analysis method for a porous blockage in a fuel subassembly is developed using the standard k-ε turbulence model with the typical correlations in handbooks. The purpose of this analysis method is to evaluate the position and the magnitude of the maximum temperature, and to investigate the thermo-hydraulic phenomena in the porous blockage. Verification of this analysis method was conducted based on the results of 4-subchannel geometry water test. It was revealed that the evaluation of the porosity distribution and the particle diameter in a porous blockage was important to predict the temperature distribution. This analysis method could simulate the spatial characteristic of velocity and temperature distributions in the blockage and evaluate the pin surface temperature inside the porous blockage. Through the verification of this analysis method, it is shown that this multi-dimensional analysis method is useful to predict the thermo-hydraulic field and the highest temperature in a porous blockage. (author)

Sodium-fuel interaction: dropping experiments and subassembly test

International Nuclear Information System (INIS)

Holtbecker, H.; Schins, H.; Jorzik, E.; Klein, K.

1978-01-01

Nine dropping tests, which bring together 2 to 4 kg of molten UO 2 with 150 l sodium, showed the incoherency and non-violence of these thermal interactions. The pressures can be described by sodium incipient boiling and bubble collapse; the UO 2 fragmentation by thermal stress and bubble collapse impact forces. The mildness of the interaction is principally due to the slowness and incoherency of UO 2 fragmentation. This means that parametric models which assume instantaneous mixing and fragmentation are of no use for the interpretation of dropping experiments. One parametric model, the Caldarola Fuel Coolant Interaction Variable Mass model, is being coupled to the two dimensional time dependent hydrodynamic REXCO-H code. In a first step the coupling is applicated to a monodimensional geometry. A subassembly test is proposed to validate the model. In this test rapid mixing between UO 2 and sodium has to be obtained. Dispersed molten UO 2 fuel is obtained by flashing injected sodium drops inside a UO 2 melt. This flashing is theoretically explained and modelled as a superheat limited explosion. The measured sodium drop dwell times of two experiments are compared to results obtained from the mentioned theory, which is the basis of the Press 2 Code

The EBR-II materials-surveillance program. 5: Results of SURV-5

International Nuclear Information System (INIS)

Ruther, W.E.; Staffon, J.D.; Carlson, B.G.; Allen, T.R.

1998-01-01

In March of 1965, a set of surveillance (SURV) samples was placed in the EBR-II reactor to determine the effect of irradiation, thermal aging, and sodium corrosion on reactor materials. Eight subassemblies were placed into row 12 positions of EBR-II to determine the effect of irradiation at 370 C. Two subassemblies were placed into the primary sodium basket to determine the effect of thermal aging at 370 C. One half of all samples were exposed to primary system sodium while one half were sealed in capsules with a helium atmosphere. Fifteen different structural materials were tested in the SURV program. In this work, the properties of these materials irradiated at 370 C to a total fluence of 3.2 x 10 22 n/cm 2 were determined. These materials are the fifth set of irradiated subassemblies to be examined as part of the SURV program (SURV-5). The properties analyzed were weight, density, microstructure, hardness, tensile and yield strength, and fracture resistance. Of all the alloys examined in SURV-5, only Berylco-25 showed any significant weight loss. Stainless steel (both 304 and 347) had the largest density decrease, although the density decrease from irradiation for all alloys was less than 0.4 percent. The microstructure of both Berylco-25 and the aluminum-bronze alloy was altered significantly. Iron- and nickel-base alloys showed little change in microstructure. Austenitic steels (304 and 347) harden with irradiation. The hardness of Inconel X750 did not change significantly with irradiation. The ultimate tensile strength of Inconel X750, 304 stainless steel, 420 stainless steel and welded 304 changed little due to a fluence increase from 2.2 x 10 22 n/cm 2 (the maximum fluence of the SURV-4 samples) to 3.2 x 10 22 n/cm 2

Specialists' meeting on thermodynamics of FBR fuel subassemblies under nominal and non-nominal operating conditions. Summary report

Energy Technology Data Exchange (ETDEWEB)

1979-01-01

The purpose of the meeting was to provide a forum for exchange of information on thermo- and fluiddynamic investigations of LMFBR-subassembly. Special emphasis was placed on nominal and non-nominal conditions. The technical part of the meeting was divided into four sessions, as follows: status of the thermo- and fluiddynamic activities; physical and mathematical modelling of single phase; rod bundle thermohydraulics; experimental investigations; and future R and D. Separate abstracts are included for each of the papers.

Design and adjustment on test bed of replacing subassembly machine control system for China experimental fast reactor

International Nuclear Information System (INIS)

Dong Shengguo; Ma Hongsheng; Zhao Lixia

2008-01-01

The present research concerns in the design and adjustment of replacing sub- assembly machine control system of China Experimental Fast Reactor. The design of replacing subassembly machine control system adopts some electric equipments, such as programmable controllers, digital DC drivers. The designed control system was adjusted on the test bed. The results indicate that the operation of the control system is steady and reliable, and designed control system can meet the needs of the design specification. (authors)

A conceptual design and structural stabilities of in-pit assembly tools for the completion of final sector assembly at tokamak hall

International Nuclear Information System (INIS)

Nam, K.O.; Park, H.K.; Kim, D.J.; Ahn, H.J.; Kim, K.K.; Im, K.; Shaw, R.

2010-01-01

The final assembly of main components of the International Thermonuclear Experimental Reactor (ITER) tokamak, Vacuum Vessel (VV) and Toroidal Field Coils (TFCs), is achieved by the sequential assembly of the nine sub-assembled 40 o sectors in tokamak pit. Each sub-assembled 40 o sector is composed of one VV 40 o sector, two TFCs, and in-between Vacuum Vessel Thermal Shield (VVTS) segments. Sub-assembly is carried out in the assembly building and then the sub-assembled sectors are transferred into tokamak pit, in sequence, to complete sector assembly. The role of in-pit assembly tool is to support and align the sub-assembled sectors in tokamak pit. It also plays the role of reference datum during assembly until the completion of main components assembly. Korea Domestic Agency (KO DA) has developed the conceptual design of most ITER purpose-built assembly tools under the collaboration with the ITER Organization. Among the conceptual designs carried out, this paper describes the function, the structure, the selected material and the design results of the in-pit assembly tools comprising central column, radial beams and their supports, TF inner supports and in-pit working floor. The results of structural analysis using ANSYS for the various loading cases are given as well. The resultant stresses and deflections turned out to fall within the allowable ranges.

Material testing facilities and programs for plasma-facing component testing

Science.gov (United States)

Linsmeier, Ch.; Unterberg, B.; Coenen, J. W.; Doerner, R. P.; Greuner, H.; Kreter, A.; Linke, J.; Maier, H.

2017-09-01

Component development for operation in a large-scale fusion device requires thorough testing and qualification for the intended operational conditions. In particular environments are necessary which are comparable to the real operation conditions, allowing at the same time for in situ/in vacuo diagnostics and flexible operation, even beyond design limits during the testing. Various electron and neutral particle devices provide the capabilities for high heat load tests, suited for material samples and components from lab-scale dimensions up to full-size parts, containing toxic materials like beryllium, and being activated by neutron irradiation. To simulate the conditions specific to a fusion plasma both at the first wall and in the divertor of fusion devices, linear plasma devices allow for a test of erosion and hydrogen isotope recycling behavior under well-defined and controlled conditions. Finally, the complex conditions in a fusion device (including the effects caused by magnetic fields) are exploited for component and material tests by exposing test mock-ups or material samples to a fusion plasma by manipulator systems. They allow for easy exchange of test pieces in a tokamak or stellarator device, without opening the vessel. Such a chain of test devices and qualification procedures is required for the development of plasma-facing components which then can be successfully operated in future fusion power devices. The various available as well as newly planned devices and test stands, together with their specific capabilities, are presented in this manuscript. Results from experimental programs on test facilities illustrate their significance for the qualification of plasma-facing materials and components. An extended set of references provides access to the current status of material and component testing capabilities in the international fusion programs.

Influence of leakage flow on the behaviour of gas behind a blockage in LMFBR subassembly geometry

International Nuclear Information System (INIS)

Fukuzawa, Y.

1980-07-01

Observations were made of the behaviour of gas behind a uniform porous 21% corner blockage within a pin-bundle of LMFBR subassembly geometry. The main parameter of the experiment was the leakage flow rate through the blockage. The behaviour of gas is significantly influenced by the leakage flow rate. The measured size and residence time of a gas cavity formed behind the blockage are shown and the mechanisms of the gas cavity dispersion by the leakage flow discussed by using a simple model of the liquid flow distribution behind the blockage. (orig.) [de

Interrelationship betwen material strength and component design under elevated temperature for FBR

International Nuclear Information System (INIS)

Nakagawa, Y.

Structural design under elevated temperature for fast breeder reactor plant is very troublesome compared to that of for lower temperature. This difficulty can be mainly discussed from two different stand points. One is design and design code, another is material strength. Components in FBR are operated under creep regime and time dependent creep behaviour should be elevated properly. This means the number and combinations of design code and material strength are significantly large and makes these systems very complicated. Material selection is, in no words, not an easy job. This should be done by not only material development but also component design stand point. With valuable experience of construction and research on FBR, a lot of information on component design and material behaviour is available. And it is a time to choose the ''best material'' from the entire stand points of component construction. (author)

Volatile components and continental material of planets

International Nuclear Information System (INIS)

Florenskiy, K.P.; Nikolayeva, O.V.

1984-01-01

It is shown that the continental material of the terrestrial planets varies in composition from planet to planet according to the abundances and composition of true volatiles (H 2 0, CO 2 , etc.) in the outer shells of the planets. The formation of these shells occurs very early in a planet's evolution when the role of endogenous processes is indistinct and continental materials are subject to melting and vaporizing in the absence of an atmosphere. As a result, the chemical properties of continental materials are related not only to fractionation processes but also to meltability and volatility. For planets retaining a certain quantity of true volatile components, the chemical transformation of continental material is characterized by a close interaction between impact melting vaporization and endogeneous geological processes

Current trends in degradation assesment on metallic materials of industrial components

International Nuclear Information System (INIS)

Herrera Palma, Victoria

2007-01-01

To needs to assess objectively a structural integrity analysis in nuclear and termal power-, oil- and chemical- industry system, represents a large challenge for engineer and researches related to Materials Science, equipment manufactures or users. These systems share many of their problems with regards to aging mechanism of components metallic materials, high replacement costs and increasing requirements on efficiency and safety. This paper makes an attempt to give an overview of the current trends on material damage and residual life assessment for installation of power-, oil- and chemical industry. Some of the currently existing ideas on components inspection, as an activity for damage detection are shown. A summary on mechanism of material damage and experimental techniques for their characterization is also presented. Finally, some analytical methods with wide appliance in materials damage evaluation and residual life assesment of components are described

Specialists' meeting on thermodynamics of FBR fuel subassemblies under nominal and non-nominal operating conditions. Summary report

Energy Technology Data Exchange (ETDEWEB)

1979-01-01

The purpose of the meeting was to provide a forum for exchange of information on thermo- and fluiddynamic investigations of LMFBR-subassembly. Special emphasis was placed on nominal and non-nominal conditions. The technical part of the meeting was divided into four sessions, as follows: status of the thermo- and fluiddynamic activities; physical and mathematical modelling of single phase; rod bundle thermohydraulics; experimental investigations; and future R and D. Separate abstracts are included for each of the papers.

Materials for 300 to 5000C magnetic components

International Nuclear Information System (INIS)

Weichold, M.H.; Pandey, R.K.; Palmer, D.W.

1980-01-01

Core materials and winding wire for audio and rf transformers have been investigated to 500 0 C. Audio cores of 2 V Permendur had parameter stability from 25 to 500 0 C and during aging at 450 0 C. High frequency ferrite material, Mix 63, displayed usefulness up to 300 0 C. Both anodized aluminum and ceramic-coated copper wire function to 500 0 C in low voltage or large gauge applications. Components based on these materials operate reliably to 500 0 C

An Assessment of the International Space Station's Trace Contaminant Control Subassembly Process Economics

Science.gov (United States)

Perry J. L.; Cole, H. E.; El-Lessy, H. N.

2005-01-01

The International Space Station (ISS) Environmental Control and Life Support System includes equipment speci.cally designed to actively remove trace chemical contamination from the cabin atmosphere. In the U.S. on-orbit segment, this function is provided by the trace contaminant control subassembly (TCCS) located in the atmosphere revitalization subsystem rack housed in the laboratory module, Destiny. The TCCS employs expendable adsorbent beds to accomplish its function leading to a potentially signi.cant life cycle cost over the life of the ISS. Because maintaining the TCCSs proper can be logistically intensive, its performance in .ight has been studied in detail to determine where savings may be achieved. Details of these studies and recommendations for improving the TCCS s process economics without compromising its performance or crew health and safety are presented and discussed.

Composition of estuarine colloidal material: organic components

Science.gov (United States)

Sigleo, A.C.; Hoering, T.C.; Helz, G.R.

1982-01-01

Colloidal material in the size range 1.2 nm to 0.4 ??m was isolated by ultrafiltration from Chesapeake Bay and Patuxent River waters (U.S.A.). Temperature controlled, stepwise pyrolysis of the freeze-dried material, followed by gas chromatographic-mass spectrometric analyses of the volatile products indicates that the primary organic components of this polymer are carbohydrates and peptides. The major pyrolysis products at the 450??C step are acetic acid, furaldehydes, furoic acid, furanmethanol, diones and lactones characteristic of carbohydrate thermal decomposition. Pyrroles, pyridines, amides and indole (protein derivatives) become more prevalent and dominate the product yield at the 600??C pyrolysis step. Olefins and saturated hydrocarbons, originating from fatty acids, are present only in minor amounts. These results are consistent with the composition of Chesapeake phytoplankton (approximately 50% protein, 30% carbohydrate, 10% lipid and 10% nucleotides by dry weight). The pyrolysis of a cultured phytoplankton and natural particulate samples produced similar oxygen and nitrogencontaining compounds, although the proportions of some components differ relative to the colloidal fraction. There were no lignin derivatives indicative of terrestrial plant detritus in any of these samples. The data suggest that aquatic microorganisms, rather than terrestrial plants, are the dominant source of colloidal organic material in these river and estuarine surface waters. ?? 1982.

Individualized FAC on bottom tab subassemblies to minimize adhesive gap between emitter and optics

Science.gov (United States)

Sauer, Sebastian; Müller, Tobias; Haag, Sebastian; Beleke, Andreas; Zontar, Daniel; Baum, Christoph; Brecher, Christian

2017-02-01

High Power Diode Laser (HPDL) systems with short focal length fast-axis collimators (FAC) require submicron assembly precision. Conventional FAC-Lens assembly processes require adhesive gaps of 50 microns or more in order to compensate for component tolerances (e.g. deviation of back focal length) and previous assembly steps. In order to control volumetric shrinkage of fast-curing UV-adhesives shrinkage compensation is mandatory. The novel approach described in this paper aims to minimize the impact of volumetric shrinkage due to the adhesive gap between HPDL edge emitters and FAC-Lens. Firstly, the FAC is actively aligned to the edge emitter without adhesives or bottom tab. The relative position and orientation of FAC to emitter are measured and stored. Consecutively, an individual subassembly of FAC and bottom tab is assembled on Fraunhofer IPT's mounting station with a precision of +/-1 micron. Translational and lateral offsets can be compensated, so that a narrow and uniform glue gap for the consecutive bonding process of bottom tab to heatsink applies (Figure 4). Accordingly, FAC and bottom tab are mounted to the heatsink without major shrinkage compensation. Fraunhofer IPT's department assembly of optical systems and automation has made several publications regarding active alignment of FAC lenses [SPIE LASE 8241-12], volumetric shrinkage compensation [SPIE LASE 9730-28] and FAC on bottom tab assembly [SPIE LASE 9727-31] in automated production environments. The approach described in this paper combines these and is the logical continuation of that work towards higher quality of HPDLs.

Materials and Components Technology Division research summary, 1991

International Nuclear Information System (INIS)

1991-04-01

This division has the purpose of providing a R and D capability for design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs are in support of the Integral Fast Reactor, life extension for light water reactors, fuels development for the new production reactor and research and test reactors, fusion reactor first-wall and blanket technology, safe shipment of hazardous materials, fluid mechanics/materials/instrumentation for fossile energy systems, and energy conservation and renewables (including tribology, high- temperature superconductivity). Separate abstracts have been prepared for the data base

Materials and Components Technology Division research summary, 1991

Energy Technology Data Exchange (ETDEWEB)

1991-04-01

This division has the purpose of providing a R and D capability for design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs are in support of the Integral Fast Reactor, life extension for light water reactors, fuels development for the new production reactor and research and test reactors, fusion reactor first-wall and blanket technology, safe shipment of hazardous materials, fluid mechanics/materials/instrumentation for fossile energy systems, and energy conservation and renewables (including tribology, high- temperature superconductivity). Separate abstracts have been prepared for the data base.

Brazing and machining of carbon based materials for plasma facing components

International Nuclear Information System (INIS)

Brossa, M.; Guerreschi, U.; Rossi, M.

1994-01-01

Carbon based materials in the recent years have often been considered and used as armour material in plasma facing components for several fusion devices, because of their low Z and good high temperature characteristics that are compatible with the operation of nuclear reactors. These materials are often connected (mechanically or by brazing) to metals, that allow the support and the cooling functions (heat sink materials). In the following the experience of Ansaldo Ricerche about the study and the manufacturing of plasma facing components and mockups is described with reference to the influence of the carbon materials in performing brazing junction with metals. It is interesting to observe how the different characteristics of the carbon materials influence the brazing process. ((orig.))

Determination of material behavior in 700 C turbine components under component and load specific conditions

Energy Technology Data Exchange (ETDEWEB)

Lueckemeyer, N.; Kirchner, H.; Kern, T.U. [Siemens AG, Muehlheim (Germany). Energy Sector; Reigl, M. [Alstom Power, Baden (Switzerland); Klenk, A.; Klein, T. [Stuttgart Univ. (Germany). MPA; Schwienheer, M.; Cui, L.; Scholz, A.; Berger, C. [Institut fuer Werkstoffkunde (IfW), Darmstadt (Germany)

2010-07-01

With global warming being one of mankind's greatest challenges, an increasing demand for electricity world-wide, and studies showing that fossil resources like coal and gas will remain a major source for electricity for the next couple of decades, research into the development of highest efficiency fossil power plants has become a top priority. Calculations for coal fired power plants have shown that CO{sub 2} emissions can be reduced by as much as 7% compared to the current state of the art equipment. It can be reached by increasing the live steam parameters to 700 C and 350bar. To achieve the desired operating hours at this temperature the application of nickel base materials is necessary for the main components such as rotors, inner casings and valves. Nowadays, the use of Nickel base alloys is common practice for selected gas turbine components. However, with steam turbine rotors being 1000mm in diameter and casings with wall thicknesses higher than 100mm the gas turbine application range and experience for nickel base alloys are well exceeded. This paper uses a basic design for a steam turbine to illustrate the core challenges in developing nickel based steam turbine components, such as casting, forging, nondestructive testing and welding. Suitable nickel based alloys have been investigated in research projects over the past years. The research results are summarized and an explanation is given as to why Alloy617 was selected for forged components and Alloy625 for cast components. This paper then focuses on the material behavior under long term and complex loading conditions and on the development of life time concepts for thick walled components made from these alloys. Due to the differences in the material behavior of nickel base alloys, the existing steel design philosophies cannot be completely adopted but rather must be carefully evaluated and modified where necessary. To do this, large test components were manufactured. Based on both standard tests

Gas turbine blades and disks. Materials and component behaviour

International Nuclear Information System (INIS)

1990-01-01

This progress report summarizes the research results obtained by the special research programme 339 in the years 1988 and 1989. Emphasis is given to the following aspects and problems: Optimisation of structure, protective coatings, connection between structure parameters and mechanical materials behaviour, tribologic materials and component behaviour, impacts of overall loads, and of stress and deformation state in the inelastic regime under mechanical and thermal load, and impacts of the manufacturing process on component behaviour, quality assurance. Eleven of the fifteen papers of the report have been separately analysed for the ENERGY database, and thirteen for the DELURA database. (orig./MM) With 191 figs., 13 tabs [de

Irradiation environment and materials behavior

International Nuclear Information System (INIS)

Ishino, Shiori

1992-01-01

Irradiation environment is unique for materials used in a nuclear energy system. Material itself as well as irradiation and environmental conditions determine the material behaviour. In this review, general directions of research and development of materials in an irradiation environment together with the role of materials science are discussed first, and then recent materials problems are described for energy systems which are already existing (LWR), under development (FBR) and to be realized in the future (CTR). Topics selected are (1) irradiation embrittlement of pressure vessel steels for LWRs, (2) high fluence performance of cladding and wrapper materials for fuel subassemblies of FBRs and (3) high fluence irradiation effects in the first wall and blanket structural materials of a fusion reactor. Several common topics in those materials issues are selected and discussed. Suggestions are made on some elements of radiation effects which might be purposely utilized in the process of preparing innovative materials. (J.P.N.) 69 refs

Soldadura laser de sub-conjuntos para estampagem (Tailored blanks)

DEFF Research Database (Denmark)

Olsen, Flemming Ove

1998-01-01

Laser welding has an increasing role in the automotive industry, namely on the sub-assemblies manufacturing. Several sheet-shape parts are laser welded, on a dissimilar combination of thicknesses and materials, and are afterwards formed (stamped) being transformed in a vehicle body component. In ...

Origin and type of flaws in heat engine ceramic materials and components

International Nuclear Information System (INIS)

Govila, R.K.

1995-01-01

A number of ceramic materials such as Silicon Nitrides and Carbides, Sialons, Whisker-Reinforced Ceramic Composites and Partially-Stabilized Zirconias (PSZs) have been developed for use as structural components in heat engine applications. The reliability and durability of a structural engine component is critically dependent on the size, density of distribution and location of flaws. This information is critical for the processing and design engineers in order to design structural components using suitable materials and thus minimize stress intensity. In general, the failure initiating flaws are associated or produced due to material impurity, processing methods and parameters, and fabrication techniques (machining and grinding). Examples of each type of flaws associated with material impurity, processing methods and fabrication techniques are illustrated

Materials for heat flux components of the first wall in fusion reactors

International Nuclear Information System (INIS)

Hoven, H.; Koizlik, K.; Linke, J.; Nickel, H.; Wallura, E.

1985-08-01

Materials of the First Wall in near-fusion plasma machines are subjected to a complex load system resulting from the plasma-wall interaction. The materials for their part also influence the plasma. Suitable materials must be available in order to ensure that the wall components achieve a sufficiently long dwell time and that their effects on the plasma remain small and controllable. The present report discusses relations between the plasma-wall interaction, the reactions of the materials and testing and examination methods for specific problems in developing and selecting suitable materials for highly stressed components on the First Wall of fusion reactors. (orig.)

Dual phase magnetic material component and method of forming

Science.gov (United States)

Dial, Laura Cerully; DiDomizio, Richard; Johnson, Francis

2017-04-25

A magnetic component having intermixed first and second regions, and a method of preparing that magnetic component are disclosed. The first region includes a magnetic phase and the second region includes a non-magnetic phase. The method includes mechanically masking pre-selected sections of a surface portion of the component by using a nitrogen stop-off material and heat-treating the component in a nitrogen-rich atmosphere at a temperature greater than about 900.degree. C. Both the first and second regions are substantially free of carbon, or contain only limited amounts of carbon; and the second region includes greater than about 0.1 weight % of nitrogen.

NABUB a non-saturated model of coolant boiling in a fast reactor sub-assembly

International Nuclear Information System (INIS)

Brook, A.J.; Mills, D.S.

1975-08-01

A theoretical model is described of sodium boiling in a fast reactor sub-assembly in which the usual assumptions of a saturated vapour are not made. Instead, vapour pressure is calculated in a perfect gas basis, which enables some allowance to be made for the possible presence of non-condensables, which may inhibit the condensation f the vapour. Indications are given of the circumstances under which such inhibition might be expected to show the most marked effects, and some sample results ontained by the code are presented. These show that the coolant voiding pattern is most sensitive to restrictions on the condensing flux in the 100 to 200w/cm 2 range. If unrestricted condensation is assumed, the results of the code are in excellent agreement with more conventional saturation models. (author)

Flexible Multibody Systems Models Using Composite Materials Components

International Nuclear Information System (INIS)

Neto, Maria Augusta; Ambr'osio, Jorge A. C.; Leal, Rog'erio Pereira

2004-01-01

The use of a multibody methodology to describe the large motion of complex systems that experience structural deformations enables to represent the complete system motion, the relative kinematics between the components involved, the deformation of the structural members and the inertia coupling between the large rigid body motion and the system elastodynamics. In this work, the flexible multibody dynamics formulations of complex models are extended to include elastic components made of composite materials, which may be laminated and anisotropic. The deformation of any structural member must be elastic and linear, when described in a coordinate frame fixed to one or more material points of its domain, regardless of the complexity of its geometry. To achieve the proposed flexible multibody formulation, a finite element model for each flexible body is used. For the beam composite material elements, the sections properties are found using an asymptotic procedure that involves a two-dimensional finite element analysis of their cross-section. The equations of motion of the flexible multibody system are solved using an augmented Lagrangian formulation and the accelerations and velocities are integrated in time using a multi-step multi-order integration algorithm based on the Gear method

A systematic concept of assuring structural integrity of components and parts for applying to highly ductile materials through brittle material

International Nuclear Information System (INIS)

Suzuki, Kazuhiko

2007-09-01

Concepts of assuring structural integrity of plant components have been developed under limited conditions of either highly ductile or brittle materials. There are some cases where operation in more and more severe conditions causes a significant reduction in ductility for materials with a high ductility before service. Use of high strength steels with relatively reduced ductility is increasing as industry applications. Current concepts of structural integrity assurance under the limited conditions of material properties or on the requirement of no significant changes in material properties even after long service will fail to incorporate expected technological innovations. A systematic concept of assuring the structural integrity should be developed for applying to highly ductile materials through brittle materials. Objectives of the on-going research are to propose a detail of the systematic concept by considering how we can develop the concept without restricting materials and for systematic considerations on a broad range of material properties from highly ductile materials through brittle materials. First, background of concepts of existing structural codes for components of highly ductile materials or for structural parts of brittle materials are discussed. Next, issues of existing code for parts of brittle materials are identified, and then resolutions to the issues are proposed. Based on the above-mentioned discussions and proposals, a systematic concept is proposed for application to components with reduced ductility materials and for applying to components of materials with significantly changing material properties due to long service. (author)

Mechanical and materials engineering of modern structure and component design

CERN Document Server

Altenbach, Holm

2015-01-01

This book presents the latest findings on mechanical and materials engineering as applied to the design of modern engineering materials and components. The contributions cover the classical fields of mechanical, civil and materials engineering, as well as bioengineering and advanced materials processing and optimization. The materials and structures discussed can be categorized into modern steels, aluminium and titanium alloys, polymers/composite materials, biological and natural materials, material hybrids and modern nano-based materials. Analytical modelling, numerical simulation, state-of-the-art design tools and advanced experimental techniques are applied to characterize the materials’ performance and to design and optimize structures in different fields of engineering applications.

Treated Coconut Coir Pith as Component of Cementitious Materials

OpenAIRE

Koňáková, Dana; Vejmelková, Eva; Čáchová, Monika; Siddique, Jamal Akhter; Polozhiy, Kirill; Reiterman, Pavel; Keppert, Martin; Černý, Robert

2015-01-01

The presented paper deals with utilization of raw and treated coir pith as potential component of cementitious composites. The studied material is coir pith originating from a coconut production. Its applicability as cement mixture component was assessed in terms of the physical properties of concrete containing different amount of coir pith. Basic physical properties, compressive and bending strength, and hygric transport characteristics as well as thermal properties belong among the studied...

Computer Simulation of Material Flow in Warm-forming Bimetallic Components

Science.gov (United States)

Kong, T. F.; Chan, L. C.; Lee, T. C.

2007-05-01

Bimetallic components take advantage of two different metals or alloys so that their applicable performance, weight and cost can be optimized. However, since each material has its own flow properties and mechanical behaviour, heterogeneous material flows will occur during the bimetal forming process. Those controls of process parameters are relatively more complicated than forming single metals. Most previous studies in bimetal forming have focused mainly on cold forming, and less relevant information about the warm forming has been provided. Indeed, changes of temperature and heat transfer between two materials are the significant factors which can highly influence the success of the process. Therefore, this paper presents a study of the material flow in warm-forming bimetallic components using finite-element (FE) simulation in order to determine the suitable process parameters for attaining the complete die filling. A watch-case-like component made of stainless steel (AISI-316L) and aluminium alloy (AL-6063) was used as the example. The warm-forming processes were simulated with the punch speeds V of 40, 80, and 120 mm/s and the initial temperatures of the stainless steel TiSS of 625, 675, 725, 775, 825, 875, 925, 975, and 1025 °C. The results showed that the AL-6063 flowed faster than the AISI-316L and so the incomplete die filling was only found in the AISI-316L region. A higher TiSS was recommended to avoid incomplete die filling. The reduction of V is also suggested because this can save the forming energy and prevent the damage of tooling. Eventually, with the experimental verification, the results from the simulation were in agreement with those of the experiments. On the basis of the results of this study, engineers can gain a better understanding of the material flow in warm-forming bimetallic components, and be able to determine more efficiently the punch speed and initial material temperature for the process.

Under sodium reliability tests on core components and in-core instrumentation

International Nuclear Information System (INIS)

Ruppert, E.; Stehle, H.; Vinzens, K.

1977-01-01

A sodium test facility for fast breeder core components (AKB), built by INTERATOM at Bensberg, has been operating since 1971 to test fuel dummies and blanket elements as well as absorber elements under simulated normal and extreme reactor conditions. Individual full-scale fuel or blanket elements and arrays of seven elements, modelling a section of the SNR-300 reactor core, have been tested under a wide range of sodium mass flow and isothermal test conditions up to 925K as well as under cyclic changed temperature transients. Besides endurance testing of the core components a special sodium and high-temperature instrumentation is provided to investigate thermohydraulic and vibrational behaviour of the test objects. During all test periods the main subassembly characteristics could be reproduced and the reliability of the instrumentation could be proven. (orig.) [de

Fundamental water experiment on subassembly with porous blockage in 4 sub-channel geometry. Influence of flow on temperature distribution in the porous blockage

International Nuclear Information System (INIS)

Tanaka, Masa-aki; Kobayashi, Jun; Isozaki, Tadasi; Nishimura, Motohiko; Kamide, Hideki

1998-03-01

In the liquid metal cooled Fast Breeder Reactor, Local Fault incident is recognized as a key issue of the local subassembly accident. In terms of the reactor safety assessment, it is important to predict the velocity and temperature distributions not only in the fuel subassembly but also in the blockage accurately to evaluate the location of the hottest point and the maximum temperature. In this study, the experiment was performed with the 4 sub-channel geometry water test facility. Dimension is five times larger than that of a real FBR. The porous blockage is located at the center sub-channel in the test section and surrounded with three unplugged sub-channels. The blockages used in this study were, the solid metal, the porous medium consisted of metal spheres, the porous blockage with end plates covering the side or top faces of the blockage to prevent the horizontal and axial flows into the blockage. The experimental parameters were the heater output provided by the electrical heater in the simulated fuel pins and the flow rate. Temperature of the fluid was measured inside/outside the blockage and velocity profiles outside the blockage were measured. (J.P.N.)

Treated Coconut Coir Pith as Component of Cementitious Materials

Directory of Open Access Journals (Sweden)

Dana Koňáková

2015-01-01

Full Text Available The presented paper deals with utilization of raw and treated coir pith as potential component of cementitious composites. The studied material is coir pith originating from a coconut production. Its applicability as cement mixture component was assessed in terms of the physical properties of concrete containing different amount of coir pith. Basic physical properties, compressive and bending strength, and hygric transport characteristics as well as thermal properties belong among the studied characteristics. It was proved that the concrete with 5% (by mass of cement of this waste material shows appropriate physical properties and it gives rise to an applicable material for building structures. Generally, the coir pith can be regarded as lightening additive. When 10% of coir pith was added, it has led to higher deterioration of properties than what is acceptable since such dosing is greatly increasing the total porosity. The influence of chemical treatment of coir pith was evaluated as well; both tested treatment methods improved the performance of cementitious composites while the acetylation was somewhat more effective the treatment by NaOH.

Development of computer code models for analysis of subassembly voiding in the LMFBR

International Nuclear Information System (INIS)

Hinkle, W.

1979-12-01

The research program discussed in this report was started in FY1979 under the combined sponsorship of the US Department of Energy (DOE), General Electric (GE) and Hanford Engineering Development Laboratory (HEDL). The objective of the program is to develop multi-dimensional computer codes which can be used for the analysis of subassembly voiding incoherence under postulated accident conditions in the LMFBR. Two codes are being developed in parallel. The first will use a two fluid (6 equation) model which is more difficult to develop but has the potential for providing a code with the utmost in flexibility and physical consistency for use in the long term. The other will use a mixture (< 6 equation) model which is less general but may be more amenable to interpretation and use of experimental data and therefore, easier to develop for use in the near term. To assure that the models developed are not design dependent, geometries and transient conditions typical of both foreign and US designs are being considered

A model of gas cavity breakup behind a blockage in fast breeder reactor subassembly geometry

International Nuclear Information System (INIS)

Fukuzawa, Y.

1980-05-01

A semi-empirical model has been developed to describe the transient behaviour of a gas cavity due to breakup behind a blockage in Liquid Metal Fast Breeder Reactor subassembly geometry. The main mechanisms assumed for gas cavity breakup in the present model are as follows: The gas cavity is broken up by the pressure fluctuation at the interface due to turbulence in the liquid. The centrifugal force on the liquid opposes breakup. The model is able to describe experimental results on the transient behaviour of a gas cavity due to breakup after the termination of gas injection. On the basis of the present model the residence time of a gas cavity behind a blockage in sodium is predicted and the dependence of the residence time on blockage size is discussed. (orig.) [de

Materials requirements for the ITER vacuum vessel and in-vessel components - approaching the construction phase

International Nuclear Information System (INIS)

Barabash, V.; Ioki, K.; Pick, M.; Girard, J.P.; Merola, M.

2007-01-01

Full text of publication follows: The ITER activities are fully devoted toward its construction. In accordance with the ITER integrated project schedule, the procurement specifications for the manufacturing of the Vacuum Vessel should be prepared by March 2008 and the procurement specifications for the in-vessel components (first wall/blanket, divertor) by 2009. To update the design, considering design and technology evolution, the ITER Design Review has been launched. Among the various topics being discussed are the important issues related to selection of materials, material procurement, and assessment of performance during operation. The main requirements related to materials for the vacuum vessel and the in-vessel components are summarized in the paper. The specific licensing requirements are to be followed for structural materials of pressure and nuclear pressure equipment components for construction of ITER. In addition, the procurements in ITER will be done mostly 'in-kind' and it is assumed that materials for these components will be produced by different Parties. However, in accordance with the regulatory requirements and quality requirements for operation, common specifications and the general rules to fulfill these requirements are to be adopted. For some ITER components (e.g. first wall, divertor high heat flux components), the ultimate qualification of the joining technologies (Be/Cu, SS/Cu, CFC/Cu, W/Cu) is under final evaluation. Successful accomplishment of the qualification program will allow to proceed with procurements of the components for ITER. The criteria for acceptance of these components and materials after manufacturing are described and the main results will be reported. Additional materials issues, which come from the on-going manufacturing R and D program, will be also described. Finally, further materials activity during the construction phase, needs for final qualification and acceptance of materials are discussed. (authors)

Evaluation of materials for heat exchanging components in advanced helium-cooled reactors

International Nuclear Information System (INIS)

Schubert, F.

1984-01-01

The qualification of metallic materials for advanced HTR applications is based on creep behaviour, fatigue properties, structural stability and corrosion resistance. A brief state of the art is provided for the materials for heat exchanging components. The experimental results are treated with respect to the importance for the design, the characteristic of time-depend materials behaviour are evaluated. Of specific interest are the possible effects of helium on the mechanical properties. Helium, which serves as primary coolant, contains traces of reactive impurities such as hydrogen, methane, carbon monoxide and water vapor. The evaluation of the HTR materials program serves as basis for structural design rules of components with operation temperatures above 800 deg C. The materials mechanical topics are discussed. Alloy improvement and the progress in development of new alloys are reviewed. (author)

Experience on sodium removal from various components

Energy Technology Data Exchange (ETDEWEB)

Kamei, M; Kanbe, M; Yagisawa, H; Sasaki, S; Kataoka, H; Fukada, T; Ishii, Y; Saito, R; Mimoto, Y [O-arai Engineering Centre, PNC, Ibaraki-ken, Tokio (Japan)

1978-08-01

Since 1970, OEC (O-arai Engineering Center) has been Investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of 'JOYO' and Dummy fuel assembly of 'JOYO', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of 'JOYO', a sector model of Sodium-to-Air cooler of 'JOYO' and a proto-type isolation valve of 'JOYO' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental sub-assemblies, the Fuel Handling Machine of 'MONJU' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a 'JOYO' prototype pump by reinstalling it after sodium removal five times. (author)

Experience on sodium removal from various components

International Nuclear Information System (INIS)

Kamei, M.; Kanbe, M.; Yagisawa, H.; Sasaki, S.; Kataoka, H.; Fukada, T.; Ishii, Y.; Saito, R.; Mimoto, Y.

1978-01-01

Since 1970, OEC (O-arai Engineering Center) has been Investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of 'JOYO' and Dummy fuel assembly of 'JOYO', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of 'JOYO', a sector model of Sodium-to-Air cooler of 'JOYO' and a proto-type isolation valve of 'JOYO' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental sub-assemblies, the Fuel Handling Machine of 'MONJU' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a 'JOYO' prototype pump by reinstalling it after sodium removal five times. (author)

Experience on sodium removal from various components

International Nuclear Information System (INIS)

Kamei, M.; Kanbe, M.; Yagisawa, H.; Sasaki, S.; Kataoka, H.

1978-02-01

Since 1970, OEC (O-arai Engineering Center) has been investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of ''JOYO'' and Dummy fuel assembly of ''JOYO'', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of ''JOYO'', a sector model of Sodium-to-Air cooler of ''JOYO'' and a proto-type Isolation valve of ''JOYO'' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental subassemblies, the Fuel Handling Machine of ''MONJU'' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of Sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a ''JOYO'' prototype pump by reinstalling it after sodium removal five times. (author)

Analysis of singular interface stresses in dissimilar material joints for plasma facing components

Energy Technology Data Exchange (ETDEWEB)

You, J.H. E-mail: jeong-ha.you@ipp.mpg.de; Bolt, H

2001-10-01

Duplex joint structures are typical material combinations for the actively cooled plasma facing components of fusion devices. The structural integrity under the incident heat loads from the plasma is one of the most crucial issues in the technology of these components. The most critical domain in a duplex joint component is the free surface edge of the bond interface between heterogeneous materials. This is due to the fact that the thermal stress usually shows a singular intensification in this region. If the plasma facing armour tile consists of a brittle material, the existence of the stress singularity can be a direct cause of failure. The present work introduces a comprehensive analytical tool to estimate the impact of the stress singularity for duplex PFC design and quantifies the relative stress intensification in various materials joints by use of a model formulated by Munz and Yang. Several candidate material combinations of plasma facing armour and metallic heat sink are analysed and the results are compared with each other.

Analysis of singular interface stresses in dissimilar material joints for plasma facing components

International Nuclear Information System (INIS)

You, J.H.; Bolt, H.

2001-01-01

Duplex joint structures are typical material combinations for the actively cooled plasma facing components of fusion devices. The structural integrity under the incident heat loads from the plasma is one of the most crucial issues in the technology of these components. The most critical domain in a duplex joint component is the free surface edge of the bond interface between heterogeneous materials. This is due to the fact that the thermal stress usually shows a singular intensification in this region. If the plasma facing armour tile consists of a brittle material, the existence of the stress singularity can be a direct cause of failure. The present work introduces a comprehensive analytical tool to estimate the impact of the stress singularity for duplex PFC design and quantifies the relative stress intensification in various materials joints by use of a model formulated by Munz and Yang. Several candidate material combinations of plasma facing armour and metallic heat sink are analysed and the results are compared with each other

Measurement of residual stresses in deposited films of SOFC component materials

Energy Technology Data Exchange (ETDEWEB)

Kato, T.; Momma, A.; Nagata, S.; Kasuga, Y. [Electrotechnical Lab., Ibaraki (Japan)

1996-12-31

The stress induced in Solid oxide fuel cells (SOFC)s has important influence on the lifetime of SOFC. But the data on stress in SOFC and mechanical properties of SOW component materials have not been accumulated enough to manufacture SOFC. Especially, the data of La{sub 1-x}Sr{sub x}MnO{sub 3} cathode and La{sub 1-x}Sr{sub x}CrO{sub 3} interconnection have been extremely limited. We have estimated numerically the dependences of residual stress in SOFC on the material properties, the cell structure and the fabrication temperatures of the components, but these unknown factors have caused obstruction to simulate the accurate behavior of residual stress. Therefore, the residual stresses in deposited La{sub 1-x}Sr{sub x}MnO{sub 3} and La{sub 1-x}Sr{sub x}CrO{sub 3} films are researched by the observation of the bending behavior of the substrate strips. The films of SOFC component materials were prepared by the RF sputtering method, because: (1) It can fabricate dense films of poor sinterable material such as La{sub 1-x}Sr{sub x}CrO{sub 3} compared with sintering or plasma spray method. (2) For the complicated material such as perovskite materials, the difference between the composition of a film and that of a target material is generally small. (3) It can fabricate a thick ceramics film by improving of the deposition rate. For example, Al{sub 2}O{sub 3} thick films of 50{mu}m can be fabricated with the deposition rate of approximately 5{mu}m/h industrially. In this paper, the dependence of residual stress on the deposition conditions is defined and mechanical properties of these materials are estimated from the results of the experiments.

Technical meeting on materials for in-vessel components of ITER

International Nuclear Information System (INIS)

Kalinin, G.; Barabash, V.

2000-01-01

The Technical meeting on materials for in-vessel components of ITER was held at the ITER Joint Work Site in Garching from 31 January to 4 February. The main objectives of the meetings were: 1. to summarize the requirements, 2. to review new data, 3. to discuss in detail the R and D program and to discuss the material assessment report

Advanced Materials Test Methods for Improved Life Prediction of Turbine Engine Components

National Research Council Canada - National Science Library

Stubbs, Jack

2000-01-01

Phase I final report developed under SBIR contract for Topic # AF00-149, "Durability of Turbine Engine Materials/Advanced Material Test Methods for Improved Use Prediction of Turbine Engine Components...

National Ignition Facility quality assurance plan for laser materials and optical technology

Energy Technology Data Exchange (ETDEWEB)

Wolfe, C.R.

1996-05-01

Quality achievement is the responsibility of the line organizations of the National Ignition Facility (NIF) Project. This subtier Quality Assurance Plan (QAP) applies to activities of the Laser Materials & Optical Technology (LM&OT) organization and its subcontractors. It responds to the NIF Quality Assurance Program Plan (QAPP, L-15958-2, NIF-95-499) and Department of Energy (DOE) Order 5700.6C. This Plan is organized according to 10 Quality Assurance (QA) criteria and subelements of a management system as outlined in the NIF QAPP. This Plan describes how those QA requirements are met. This Plan is authorized by the Associate Project Leader for the LM&OT organization, who has assigned responsibility to the Optics QA engineer to maintain this plan, with the assistance of the NIF QA organization. This Plan governs quality-affecting activities associated with: design; procurement; fabrication; testing and acceptance; handling and storage; and installation of NIF Project optical components into mounts and subassemblies.

National Ignition Facility quality assurance plan for laser materials and optical technology

International Nuclear Information System (INIS)

Wolfe, C.R.

1996-05-01

Quality achievement is the responsibility of the line organizations of the National Ignition Facility (NIF) Project. This subtier Quality Assurance Plan (QAP) applies to activities of the Laser Materials ampersand Optical Technology (LM ampersand OT) organization and its subcontractors. It responds to the NIF Quality Assurance Program Plan (QAPP, L-15958-2, NIF-95-499) and Department of Energy (DOE) Order 5700.6C. This Plan is organized according to 10 Quality Assurance (QA) criteria and subelements of a management system as outlined in the NIF QAPP. This Plan describes how those QA requirements are met. This Plan is authorized by the Associate Project Leader for the LM ampersand OT organization, who has assigned responsibility to the Optics QA engineer to maintain this plan, with the assistance of the NIF QA organization. This Plan governs quality-affecting activities associated with: design; procurement; fabrication; testing and acceptance; handling and storage; and installation of NIF Project optical components into mounts and subassemblies

Armour Materials for the ITER Plasma Facing Components

Science.gov (United States)

Barabash, V.; Federici, G.; Matera, R.; Raffray, A. R.; ITER Home Teams,

The selection of the armour materials for the Plasma Facing Components (PFCs) of the International Thermonuclear Experimental Reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-à-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R&D.

Armour materials for the ITER plasma facing components

International Nuclear Information System (INIS)

Barabash, V.; Federici, G.; Matera, R.; Raffray, A.R.

1999-01-01

The selection of the armour materials for the plasma facing components (PFCs) of the international thermonuclear experimental reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-a-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R and D. (orig.)

Reactor materials program process water component failure probability

International Nuclear Information System (INIS)

Daugherty, W. L.

1988-01-01

The maximum rate loss of coolant accident for the Savannah River Production Reactors is presently specified as the abrupt double-ended guillotine break (DEGB) of a large process water pipe. This accident is not considered credible in light of the low applied stresses and the inherent ductility of the piping materials. The Reactor Materials Program was initiated to provide the technical basis for an alternate, credible maximum rate LOCA. The major thrust of this program is to develop an alternate worst case accident scenario by deterministic means. In addition, the probability of a DEGB is also being determined; to show that in addition to being mechanistically incredible, it is also highly improbable. The probability of a DEGB of the process water piping is evaluated in two parts: failure by direct means, and indirectly-induced failure. These two areas have been discussed in other reports. In addition, the frequency of a large bread (equivalent to a DEGB) in other process water system components is assessed. This report reviews the large break frequency for each component as well as the overall large break frequency for the reactor system

Specification of steam generator, condenser and regenerative heat exchanger materials for nuclear applications

International Nuclear Information System (INIS)

Jovasevic, J.V.; Stefanovic, V.M.; Spasic, Z.LJ.

1977-01-01

The basic standards specifications of materials for nuclear applications are selected. Seamless Ni-Cr-Fe alloy Tubes (Inconel-600) for steam generators, condensers and other heat exchangers can be employed instead of austenitic stainless steal or copper alloys tubes; supplementary requirements for these materials are given. Specifications of Ni-Cr-Fe alloy plate, sheet and strip for steam generator lower sub-assembly, U-bend seamless copper-alloy tubes for heat exchanger and condensers are also presented. At the end, steam generator channel head material is proposed in the specification for carbon-steel castings suitable for welding

Low cost, small form factor, and integration as the key features for the optical component industry takeoff

Science.gov (United States)

Schiattone, Francesco; Bonino, Stefano; Gobbi, Luigi; Groppi, Angelamaria; Marazzi, Marco; Musio, Maurizio

2003-04-01

In the past the optical component market has been mainly driven by performances. Today, as the number of competitors has drastically increased, the system integrators have a wide range of possible suppliers and solutions giving them the possibility to be more focused on cost and also on footprint reduction. So, if performances are still essential, low cost and Small Form Factor issues are becoming more and more crucial in selecting components. Another evolution in the market is the current request of the optical system companies to simplify the supply chain in order to reduce the assembling and testing steps at system level. This corresponds to a growing demand in providing subassemblies, modules or hybrid integrated components: that means also Integration will be an issue in which all the optical component companies will compete to gain market shares. As we can see looking several examples offered by electronic market, to combine low cost and SFF is a very challenging task but Integration can help in achieving both features. In this work we present how these issues could be approached giving examples of some advanced solutions applied to LiNbO3 modulators. In particular we describe the progress made on automation, new materials and low cost fabrication methods for the parts. We also introduce an approach in integrating optical and electrical functionality on LiNbO3 modulators including RF driver, bias control loop, attenuator and photodiode integrated in a single device.

Summary and implications of out-of-pile investigations of local cooling disturbances in LMFBR subassembly geometry under single-phase and boiling conditions

International Nuclear Information System (INIS)

Huber, F.; Peppler, W.

1985-05-01

The consequences of local cooling disturbances in subassemblies of LMFBRs have been investigated out-of-pile at KfK. Flow and temperature distributions in the disturbed region as well as cooling under boiling conditions up to loss of cooling were investigated. Fission gas release was simulated by gas injection. A total of 16 different blockages in 20 test set-ups were used, four of them under sodium and the rest under water conditions. Mainly planar plates of different sizes and arrangements were used as blockages. In some of the experiments performed in water also porous blockages were investigated. The test sections consisted of electrically heated pin bundles with a thermal-hydraulic characteristic corresponding to that of an SNR 300 subassembly. With different parameter settings the single-phase tests in water furnished a multitude of test results on flow and temperature fields and on the behaviour of gas in the recirculation zone. In the experiments involving boiling two boiling patterns were observed: steady-state boiling and oscillating boiling. With increasing boiling intensity the boiling region grew to some extent, but it remained always confined to the blocked zone because of the relatively cold sodium flow around this zone. In the experiments simulating fission gas release it was found that under certain conditions gas accumulates in the reverse flow region behind a blockage and leads to loss of cooling. (orig./GL) [de

Drying and wetting of building materials and components

CERN Document Server

2014-01-01

This book, Drying and Wetting of Building Materials and Components, provides a collection of recent contributions in the field of drying and wetting in porous building materials. The main benefit of the book is that it discusses some of the most important topics related to the drying and wetting processes, namely, innovations and trends in drying science and technology, drying mechanism and theory, equipment, advanced modelling, complex simulation and experimentation. At the same time, these topics will be going to the encounter of a variety of scientific and engineering disciplines. The book is divided in several chapters that intend to be a resume of the current state of knowledge for benefit of professional colleagues.

A content analysis of preconception health education materials: characteristics, strategies, and clinical-behavioral components.

Science.gov (United States)

Levis, Denise M; Westbrook, Kyresa

2013-01-01

Many health organizations and practitioners in the United States promote preconception health (PCH) to consumers. However, summaries and evaluations of PCH promotional activities are limited. We conducted a content analysis of PCH health education materials collected from local-, state-, national-, and federal-level partners by using an existing database of partners, outreach to maternal and child health organizations, and a snowball sampling technique. Not applicable. Not applicable. Thirty-two materials were included for analysis, based on inclusion/exclusion criteria. A codebook guided coding of materials' characteristics (type, authorship, language, cost), use of marketing and behavioral strategies to reach the target population (target audience, message framing, call to action), and inclusion of PCH subject matter (clinical-behavioral components). The self-assessment of PCH behaviors was the most common material (28%) to appear in the sample. Most materials broadly targeted women, and there was a near-equal distribution in targeting by pregnancy planning status segments (planners and nonplanners). "Practicing PCH benefits the baby's health" was the most common message frame used. Materials contained a wide range of clinical-behavioral components. Strategic targeting of subgroups of consumers is an important but overlooked strategy. More research is needed around PCH components, in terms of packaging and increasing motivation, which could guide use and placement of clinical-behavioral components within promotional materials.

Packaging strategies for printed circuit board components. Volume I, materials & thermal stresses.

Energy Technology Data Exchange (ETDEWEB)

Neilsen, Michael K. (Kansas City Plant, Kansas City, MO); Austin, Kevin N.; Adolf, Douglas Brian; Spangler, Scott W.; Neidigk, Matthew Aaron; Chambers, Robert S.

2011-09-01

Decisions on material selections for electronics packaging can be quite complicated by the need to balance the criteria to withstand severe impacts yet survive deep thermal cycles intact. Many times, material choices are based on historical precedence perhaps ignorant of whether those initial choices were carefully investigated or whether the requirements on the new component match those of previous units. The goal of this program focuses on developing both increased intuition for generic packaging guidelines and computational methodologies for optimizing packaging in specific components. Initial efforts centered on characterization of classes of materials common to packaging strategies and computational analyses of stresses generated during thermal cycling to identify strengths and weaknesses of various material choices. Future studies will analyze the same example problems incorporating the effects of curing stresses as needed and analyzing dynamic loadings to compare trends with the quasi-static conclusions.

Observations of the behaviour of gas in the wake behind a corner blockage in fast breeder reactor subassembly geometry

International Nuclear Information System (INIS)

Fukuzawa, Y.

1979-07-01

Observations were made of gas behaviour in the wake behind a 21% corner blockage in the subassembly geometry of a liquid metal fast breeder reactor. The test section used represented one half of the reactor fuel subassembly, divided along the vertical plane of symmetry through the blockage. A glass wall occupied the position of this plane. Water was allowed to flow between glass rods simulating fuel pins, the velocity being changed from 1.2 to 4.5 m/s. Argon was injected into the wake or into the flow upstream of the blockage, the injection rate being changed from 1 to 230 Ncm 3 /s (standard temperature and pressure). From the present experiment, the following is evident: The gas is accumulated in the wake behind the blockage, forming a gas cavity. The flow patterns of the two-phase mixture in the wake are classified into three types, depending on the liquid velocity. In the lower velocity range, a gas cavity cannot be present at rest, rising up through the wake as a single bubble due to buoyancy. In the higher velocity range, the gas cavity is broken up by the liquid flow forces, only small gas bubbles circulating in the wake. In the velocity range in between, the gas cavity is present in the wake. The cavity size depends on the gas injection rate and on the liquid velocity. From the results, the possibility of fuel failure caused by fission gas release at a blockage in the fast breeder reactor can be considered to depend on the operating conditions of the reactor, specially on the coolant velocity. (orig.) [de

Development of process route for production of tubing for various core sub-assemblies and heat exchangers for 500 MWe Indian PFBR

International Nuclear Information System (INIS)

Lakshminarayana, B.; Phani Babu, C.; Dubey, A.K.; Surender, A.; Deshpande, K.V.K.; Maity, P.K.

2009-01-01

India's three stage Nuclear Power Program has entered its second stage on commercial scale with the commencement of construction of 500 MWe Prototype Fast Breeder Reactor (PFBR) at Kalpakkam. Nuclear Fuel Complex (NFC), Hyderabad is playing a crucial role in the manufacture of all the critical sub-assemblies and control elements for this reactor. The challenging task of process development and production of the various critical tubing for these sub assemblies for PFBR has been taken up by Stainless Steel Tubes Plant (SSTP), NFC with indigenous development of the equipment and technology

Method of Manufacturing A Porous Polymer Component Involving Use of A Dissolvable, Sacrificial Material

DEFF Research Database (Denmark)

2015-01-01

and thereby the resulting inner structure of the component 1 is arranged in a controlled and reproducible manner. The sacrificial material 2 and possibly also the component material 3 may e.g. be arranged by use of a 3D-printer or manually. The method may e.g. be used to manufacture a three...

Countermeasure technologies against materials deterioration of nuclear power plant components

International Nuclear Information System (INIS)

2004-09-01

This report was tentative safety standard on countermeasure technologies against materials deterioration of nuclear power plant components issued in 2004 on the base of the testing data obtained until March 2004, which was to be applied for technical evaluation for lifetime management of aged plants and preventive maintenance or repair of neutron irradiated components such as core shrouds and jet pumps. In order to prevent stress corrosion cracks (SCCs) of austenitic stainless steel welds of reactor components, thermal surface modification using laser beams was used on neutron irradiated materials with laser cladding or surface melting process methods by limiting heat input according to amount of accumulated helium so as to prevent crack initiation caused by helium bubble growth and coalescence. Laser cladding method of laser welding using molten sleeve set inside pipe surface to prevent SCCs of nickel-chromium-iron alloy welds, alloy 690 cladding method using tungsten inert gas (TIG) welding to prevent SCCs of nickel-chromium-iron alloy welds for dissimilar joints of pipes, and laser surface solid solution heat treatment method of laser irradiation on surfaces to prevent SCCs of austenitic stainless steel welds were also included as repair technologies. (T. Tanaka)

Liquid metal cooled fast breeder nuclear reactors

International Nuclear Information System (INIS)

Duncombe, E.; Thatcher, G.

1979-01-01

The invention described relates to a liquid metal cooled fast breeder nuclear reactor in which the fuel assembly has an inner zone comprised mainly of fissile material and a surrounding outer zone comprised mainly of breeder material. According to the invention the sub-assemblies in the outer zone include electro-magnetic braking devices (magnets, pole pieces and armature) for regulating the flow of coolant through the sub-assemblies. The magnetic fields of the electro-magnetic breaking devices are temperature sensitive so that as the power output of the breeder sub-assemblies increases the electro-magnetic resistance to coolant flow is reduced thereby maintaining the temperature of the coolant outlets from the sub-assemblies substantially constant. (UK)

Component-Level Selection and Qualification for the Global Ecosystem Dynamics Investigation (GEDI) Laser Altimeter Transmitter

Science.gov (United States)

Frese, Erich A.; Chiragh, Furqan L.; Switzer, Robert; Vasilyev, Aleksey A.; Thomes, Joe; Coyle, D. Barry; Stysley, Paul R.

2018-01-01

Flight quality solid-state lasers require a unique and extensive set of testing and qualification processes, both at the system and component levels to insure the laser's promised performance. As important as the overall laser transmitter design is, the quality and performance of individual subassemblies, optics, and electro-optics dictate the final laser unit's quality. The Global Ecosystem Dynamics Investigation (GEDI) laser transmitters employ all the usual components typical for a diode-pumped, solid-state laser, yet must each go through their own individual process of specification, modeling, performance demonstration, inspection, and destructive testing. These qualification processes and results for the laser crystals, laser diode arrays, electro-optics, and optics, will be reviewed as well as the relevant critical issues encountered, prior to their installation in the GEDI flight laser units.

Materials for Advanced Ultra-supercritical (A-USC) Steam Turbines – A-USC Component Demonstration

Energy Technology Data Exchange (ETDEWEB)

Purgert, Robert [Energy Industries Of Ohio Inc., Independence, OH (United States); Phillips, Jeffrey [Energy Industries Of Ohio Inc., Independence, OH (United States); Hendrix, Howard [Energy Industries Of Ohio Inc., Independence, OH (United States); Shingledecker, John [Energy Industries Of Ohio Inc., Independence, OH (United States); Tanzosh, James [Energy Industries Of Ohio Inc., Independence, OH (United States)

2016-10-01

The work by the United States Department of Energy (U.S. DOE)/Ohio Coal Development Office (OCDO) advanced ultra-supercritical (A-USC) Steam Boiler and Turbine Materials Consortia from 2001 through September 2015 was primarily focused on lab scale and pilot scale materials testing. This testing included air- or steam-cooled “loops” that were inserted into existing utility boilers to gain exposure of these materials to realistic conditions of high temperature and corrosion due to the constituents in the coal. Successful research and development resulted in metallic alloy materials and fabrication processes suited for power generation applications with metal temperatures up to approximately 1472°F (800°C). These materials or alloys have shown, in extensive laboratory tests and shop fabrication studies, to have excellent applicability for high-efficiency low CO₂ transformational power generation technologies previously mentioned. However, as valuable as these material loops have been for obtaining information, their scale is significantly below that required to minimize the risk associated with a power company building a multi-billion dollar A-USC power plant. To decrease the identified risk barriers to full-scale implementation of these advanced materials, the U.S. DOE/OCDO A-USC Steam Boiler and Turbine Materials Consortia identified the key areas of the technology that need to be tested at a larger scale. Based upon the recommendations and outcome of a Consortia-sponsored workshop with the U.S.’s leading utilities, a Component Test (ComTest) Program for A-USC was proposed. The A-USC ComTest program would define materials performance requirements, plan for overall advanced system integration, design critical component tests, fabricate components for testing from advanced materials, and carry out the tests. The AUSC Component Test was premised on the program occurring at multiple facilities, with the operating temperatures, pressure and/or size of

PFMC14. 14th international conference on plasma-facing materials and components for fusion applications. Book of abstracts

International Nuclear Information System (INIS)

2013-01-01

The performance of fusion devices and of a future fusion power plant critically depends on the plasma facing materials and components. Resistance to local heat and particle loads, thermo-mechanical properties, as well as the response to neutron damage of the selected materials are critical parameters which need to be understood and tailored from atomistic to component levels. The 14th International Conference on Plasma-Facing Materials and Components for Fusion Applications addresses these issues. Among the topics of the joint conference recent developments and research results in the following fields are addressed: - Tungsten and tungsten alloys - Low-Z materials - Mixed materials - Erosion, redeposition and fuel retention - Materials under extreme thermal loads - Technology and testing of plasma-facing components - Neutron effects in plasma-facing materials - Advanced characterization of materials and components. Selected international speakers present overview lectures and treat detailed aspects of the given topics. Contributed papers to the subjects of the meeting are solicited for oral and poster presentations.

Fission product concentration evolution in sodium pool following a fuel subassembly failure in an LMFBR

International Nuclear Information System (INIS)

Natesan, K.; Velusamy, K.; Selvaraj, P.; Kasinathan, N.; Chellapandi, P.; Chetal, S.; Bhoje, S.

2003-01-01

During a fuel element failure in a liquid metal cooled fast breeder reactor, the fission products originating from the failed pins mix into the sodium pool. Delayed Neutron Detectors (DND) are provided in the sodium pool to detect such failures by way of detection of delayed neutrons emitted by the fission products. The transient evolution of fission product concentration is governed by the sodium flow distribution in the pool. Transient hydraulic analysis has been carried out using the CFD code PHOENICS to estimate fission product concentration evolution in hot pool. k- ε turbulence model and zero laminar diffusivity for the fission product concentration have been considered in the analysis. Times at which the failures of various fuel subassemblies (SA) are detected by the DND are obtained. It has been found that in order to effectively detect the failure of every fuel SA, a minimum of 8 DND in hot pool are essential

Experience with the source evaluation board method of procuring technical components for the Fermilab Main Injector

International Nuclear Information System (INIS)

Harding, D.J.; Collins, J.P.; Kobliska, G.R.; Chester, N.S.; Pewitt, E.G.; Fowler, W.B.

1993-01-01

Fermilab has adopted the Source Evaluation Board (SEB) method for procuring certain major technical components of the Fermilab Main Injector. The SEB procedure is designed to ensure the efficient and effective expenditure of Government funds at the same time that it optimizes the opportunity for attainment of project objectives. A qualitative trade-off is allowed between price and technical factors. The process involves a large amount of work and is only justified for a very limited number of procurements. Fermilab has gained experience with the SEB process in awarding subcontracts for major subassemblies of the Fermilab Main Injector dipoles

Performance of materials in the component cooling water systems of pressurized water reactors

International Nuclear Information System (INIS)

Lee, B.S.

1993-01-01

The component cooling water (CCW) system provides cooling water to several important loads throughout the plant under all operating conditions. An aging assessment CCW systems in pressurized water reactors (PWRs) was conducted as part of Nuclear Plant Aging Research Program (NPAR) instituted by the US Nuclear Regulatory Commission. This paper presents some of the results on the performances of materials in respect of their application in CCW Systems. All the CCW system failures reported to the Nuclear Plant Reliability Data System (NPRDS) from January 1988 to June 1990 were reviewed; it is concluded that three of the main contributors to CCW system failures are valves, pumps, and heat exchangers. This study identified the modes and causes of failure for these components; most of the causes for the aging-related failures could be related to the performance of materials. Also, in this paper the materials used for these components are reviewed, and there aging mechanisms under CCW system conditions are discussed

Criticality safety assessment of FBTR fuel sub-assemblies using WIMS cross section set

International Nuclear Information System (INIS)

Gupta, H.C.; Chakraborty, B.

2002-01-01

Full text: FBTR's irradiated fuel sub-assemblies (FSAs) are sent to RML at Indira Gandhi Centre for Atomic Research for post irradiation examination. The FSAs are cut open and the fuel pins are separated for examination in the hot cells. It was required to evaluate the criticality safety in handling the FSAs in the hot cells. Criticality safety studies for handling two as well as three irradiated FSAs in the hot cells under dry conditions were carried out by the Safety Group at IGCAR, Kalpakkam. Monte Carlo code KENO (Version Va) which uses 16-group Hansen-Roach cross-section set was used for the calculations. Subsequently, during the safety review of the proposition by the Safety Review Committee (SARCOP) of AERB, it was stipulated to carry out the criticality safety studies under flooded condition also. We carried out the criticality safety studies for these fuel sub assemblies in different configurations under dry (buried in concrete) as well as wet condition (flooded with light water) using Monte Carlo codes MONALI (developed at BARC) and KENO4 using WlMS-69 group cross section set. Results of our analyses under various conditions are presented in this paper

Near term and long term materials issues and development needs for plasma interactive components

International Nuclear Information System (INIS)

Mattas, R.F.

1986-01-01

Plasma interactive components (PICs), including the first wall, limiter blades, divertor collector plates, halo scrapers, and RF launchers, are exposed to high particle fluxes that can result in high sputtering erosion rates and high heat fluxes. In addition, the materials in reactors are exposed to high neutron fluxes which will degrade the bulk properties. This severe environment will limit the materials and designs which can be used in fusion devices. In order to provide a reasonable degree of confidence that plasma interactive components will operate successfully, a comprehensive development program is needed. Materials research and development plays a key role in the successful development of PICs. The range of operating conditions along with a summary of the major issues for materials development is described. The areas covered include plasma/materials interactions, erosion/redeposition, baseline materials properties, fabrication, and irradiation damage effects. Candidate materials and materials development needs in the near term and long term are identified

Developing standard performance testing procedures for material control and accounting components at a site

International Nuclear Information System (INIS)

Scherer, Carolynn P.; Bushlya, Anatoly V.; Efimenko, Vladimir F.; Ilyanstev, Anatoly; Regoushevsky, Victor I.

2010-01-01

The condition of a nuclear material control and accountability system (MC and A) and its individual components, as with any system combining technical elements and documentation, may be characterized through an aggregate of values for the various parameters that determine the system's ability to perform. The MC and A system's status may be functioning effectively, marginally or not functioning based on a summary of the values of the individual parameters. This work included a review of the following subsystems, MC and A and Detecting Material Losses, and their respective elements for the material control and accountability system: (a) Elements of the MC and A Subsystem - Information subsystem (Accountancy/Inventory), Measurement subsystem, Nuclear Material Access subsystem, including tamper-indicating device (TID) program, and Automated Information-gathering subsystem; (b) Elements for Detecting Nuclear Material Loses Subsystem - Inventory Differences, Shipper/receiver Differences, Confirmatory Measurements and differences with accounting data, and TID or Seal Violations. In order to detect the absence or loss of nuclear material there must be appropriate interactions among the elements and their respective subsystems from the list above. Additionally this work includes a review of regulatory requirements for the MC and A system component characteristics and criteria that support the evaluation of the performance of the listed components. The listed components had performance testing algorithms and procedures developed that took into consideration the regulatory criteria. The developed MC and A performance-testing procedures were the basis for a Guide for MC and A Performance Testing at the material balance areas (MBAs) of State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering (SSC RF-IPPE).

IAEA consultants' meeting on thermal response of plasma facing materials and components

International Nuclear Information System (INIS)

Janev, R.K.

1990-07-01

The present Summary Report contains brief proceedings and the main conclusions and recommendations of the IAEA Consultants' Meeting on ''Thermal Response of Plasma Facing Materials and Components'', which was organized by the IAEA Atomic and Molecular Data Unit and held on June 11-13, 1990, in Vienna, Austria. The Report also includes a categorization and assessment of currently studied plasma facing materials, a classification scheme of material properties data, required in fusion reactor design, and a survey of the urgently needed material properties data. (author)

Recent developments in turbomachinery component materials and manufacturing challenges for aero engine applications

Science.gov (United States)

Srinivas, G.; Raghunandana, K.; Satish Shenoy, B.

2018-02-01

In the recent years the development of turbomachinery materials performance enhancement plays a vital role especially in aircraft air breathing engines like turbojet engine, turboprop engine, turboshaft engine and turbofan engines. Especially the transonic flow engines required highly sophisticated materials where it can sustain the entire thrust which can create by the engine. The main objective of this paper is to give an overview of the present cost-effective and technological capabilities process for turbomachinery component materials. Especially the main focus is given to study the Electro physical, Photonic additive removal process and Electro chemical process for turbomachinery parts manufacture. The aeronautical propulsion based technologies are reviewed thoroughly where in surface reliability, geometrical precession, and material removal and highly strengthened composite material deposition rates usually difficult to cut dedicated steels, Titanium and Nickel based alloys. In this paper the past aeronautical and propulsion mechanical based manufacturing technologies, current sophisticated technologies and also future challenging material processing techniques are covered. The paper also focuses on the brief description of turbomachinery components of shaping process and coating in aeromechanical applications.

RESEARCH ON THE STUDY OF MATERIAL DEFECTS AND SOMECOAL MILLS SUBASSEMBLIES LIFE TIME

Directory of Open Access Journals (Sweden)

Cristina LAPADUSI

2013-05-01

Full Text Available The defectsfrom the structureof metallic materials of whichare manufactured the pieces, canbeputoutbyNDT. One ofNDTmethods, commonly usedin practiceisultrasonicmethod.In this paper are rendered the results of the determinations by the effects of coal mills bars by type DGS 100,obtained with ultrasound devices by type PHASOR XS.

Printable Materials for the Realization of High Performance RF Components: Challenges and Opportunities

Directory of Open Access Journals (Sweden)

Eva S. Rosker

2018-01-01

Full Text Available Printing methods such as additive manufacturing (AM and direct writing (DW for radio frequency (RF components including antennas, filters, transmission lines, and interconnects have recently garnered much attention due to the ease of use, efficiency, and low-cost benefits of the AM/DW tools readily available. The quality and performance of these printed components often do not align with their simulated counterparts due to losses associated with the base materials, surface roughness, and print resolution. These drawbacks preclude the community from realizing printed low loss RF components comparable to those fabricated with traditional subtractive manufacturing techniques. This review discusses the challenges facing low loss RF components, which has mostly been material limited by the robustness of the metal and the availability of AM-compatible dielectrics. We summarize the effective printing methods, review ink formulation, and the postprint processing steps necessary for targeted RF properties. We then detail the structure-property relationships critical to obtaining enhanced conductivities necessary for printed RF passive components. Finally, we give examples of demonstrations for various types of printed RF components and provide an outlook on future areas of research that will require multidisciplinary teams from chemists to RF system designers to fully realize the potential for printed RF components.

The Design and Construction Process of a Test Stand for Casting the Power Steering’S Housing with the Use of the Pdcpd Material

Science.gov (United States)

Sobek, M.; Baier, A.; Grabowski, Ł.

2018-01-01

The use of new technologies and materials in various industries is a natural process that is directly related to the very high rate of development of these technologies. Certain industries decide to much faster introduce new technologies and materials. One of such branches is the automotive industry, whose representatives are very energetically looking for both financial savings and savings resulting from the vehicles mass reduction. An economically justified approach to construction materials is leading the search for new solutions and materials. The use of a modern material such as the two-component PDCPD composite shows hitherto unknown possibilities of producing subassemblies of many different constructions. The possibility of using a modern composite material with parameters comparable to that of metals and significantly lighter, can be an excellent alternative in the selection of materials for many parts of motor vehicles. The potentiality of precise casting of tolerated surfaces will allow to reduce the operations related to machining process, which is an indispensable part of the production process of elements that are cast of metal. This article describes the process of designing and building a test stand for precise positioning of power steering gear components at the stage of casting their housing. The article presents the principle of operation of the test stand and the process of preparation for the casting and the cast itself will be rudely described. Due to the implementation of research as part of a research project with an industrial partner, the article will only describe some operations. This is related to the confidentiality of the project.

NUMERICAL THERMAL ANALYSIS OF A CAR BRAKING SYSTEM

Directory of Open Access Journals (Sweden)

Patryk Różyło

2017-06-01

Full Text Available The study involved performing a numerical thermal analysis of selected components in a car braking system. The primary goal of the study was to determine the regions which are the most susceptible to variations in temperature, and to determine the degree of thermal impact upon them. The analysis was performed using the Abaqus environment. The examined components of the braking system were made of materials reflecting the mechanical properties of the real subassemblies. The FEM analysis enabled determination of the distribution of temperature in the system with respect to the properties of the investigated materials and applied boundary conditions.

Progress of High Heat Flux Component Manufacture and Heat Load Experiments in China

Energy Technology Data Exchange (ETDEWEB)

Liu, X.; Lian, Y.; Xu, Z.; Chen, J.; Chen, L.; Wang, Q.; Duan, X., E-mail: xliu@swip.ac.cn [Southwestern Institute of Physics, Chengu (China); Luo, G. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Yan, Q. [University of Science and Technology Beijing, Beijing (China)

2012-09-15

Full text: High heat flux components for first wall and divertor are the key subassembly of the present fusion experiment apparatus and fusion reactors in the future. It is requested the metallurgical bonding among the plasma facing materials (PFMs), heat sink and support materials. As to PFMs, ITER grade vacuum hot pressed beryllium CN-G01 was developed in China and has been accepted as the reference material of ITER first wall. Additionally pure tungsten and tungsten alloys, as well as chemical vapor deposition (CVD) W coating are being developed for the aims of ITER divertor application and the demand of domestic fusion devices, and significant progress has been achieved. For plasma facing components (PFCs), high heat flux components used for divertor chamber are being studied according to the development program of the fusion experiment reactor of China. Two reference joining techniques of W/Cu mockups for ITER divertor chamber are being developed, one is mono-block structure by pure copper casting of tungsten surface following by hot iso-static press (HIP), and another is flat structure by brazing. The critical acceptance criteria of high heat flux components are their high heat load performance. A 60 kW Electron-beam Material testing Scenario (EMS-60) has been constructed at Southwestern Institute of Physics (SWIP),which adopts an electron beam welding gun with maximum energy of 150 keV and 150 x 150 mm{sup 2} scanning area by maximum frame rate of 30 kHz. Furthermore, an Engineering Mockup testing Scenario (EMS-400) facility with 400 kW electron-beam melting gun is under construction and will be available by the end of this year. After that, China will have the comprehensive capability of high heat load evaluation from PFMs and small-scale mockups to engineering full scale PFCs. A brazed W/CuCrZr mockup with 25 x 25 x 40 mm{sup 3} in dimension was tested at EMS-60. The heating and cooling time are 10 seconds and 15 seconds, respectively. The experiment

MISER-I: a computer code for JOYO fuel management

International Nuclear Information System (INIS)

Yamashita, Yoshioki

1976-06-01

A computer code ''MISER-I'' is for a nuclear fuel management of Japan Experimental Fast Breeder Reactor JOYO. The nuclear fuel management in JOYO can be regarded as a fuel assembly management because a handling unit of fuel in JOYO plant is a fuel subassembly (core and blanket subassembly), and so the recording of material balance in computer code is made with each subassembly. The input information into computer code is given with each subassembly for a transfer operation, or with one reactor cycle and every one month for a burn-up in reactor core. The output information of MISER-I code is the fuel assembly storage record, fuel storage weight record in each material balance subarea at any specified day, and fuel subassembly transfer history record. Change of nuclear fuel composition and weight due to a burn-up is calculated with JOYO-Monitoring Code by off-line computation system. MISER-I code is written in FORTRAN-IV language for FACOM 230-48 computer. (auth.)

Method of Joining Graphite Fibers to a Substrate

Science.gov (United States)

Beringer, Durwood M. (Inventor); Caron, Mark E. (Inventor); Taddey, Edmund P. (Inventor); Gleason, Brian P. (Inventor)

2014-01-01

A method of assembling a metallic-graphite structure includes forming a wetted graphite subassembly by arranging one or more layers of graphite fiber material including a plurality of graphite fibers and applying a layer of metallization material to ends of the plurality of graphite fibers. At least one metallic substrate is secured to the wetted graphite subassembly via the layer of metallization material.

Development and verification test of integral reactor major components

Energy Technology Data Exchange (ETDEWEB)

Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

1999-03-01

The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability.

Development and verification test of integral reactor major components

International Nuclear Information System (INIS)

Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

1999-03-01

The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability

Evaluation of material integrity on electricity power steam generator cycles (turbine casing) component

International Nuclear Information System (INIS)

Histori; Benedicta; Farokhi; S A, Soedardjo; Triyadi, Ari; Natsir, M

1999-01-01

The evaluation of material integrity on power steam generator cycles component was done. The test was carried out on casing turbine which is made from Inconel 617. The tested material was taken from t anjung Priok plant . The evaluation was done by metallography analysis using microscope with magnification of 400. From the result, it is shown that the material grains are equiaxed

Research with neutron and synchrotron radiation on aerospace and automotive materials and components

Energy Technology Data Exchange (ETDEWEB)

Kaysser, Wolfgang; Abetz, Volker; Huber, Norbert; Kainer, Karl Ulrich; Pyczak, Florian; Schreyer, Andreas; Staron, Peter [Helmholtz-Zentrum Geesthacht Zentrum fuer Material und Kuestenforschung, Geesthacht (Germany); Esslinger, Joerg [MTU Aero Engines GmbH, Muenchen (Germany); Klassen, Thomas [Helmholtz-Zentrum Geesthacht Zentrum fuer Material und Kuestenforschung, Geesthacht (Germany); Helmut Schmidt Universitaet, Hamburg (Germany)

2011-08-15

Characterization with neutrons and synchrotron radiation has yielded essential contributions to the research and development of automotive and aerospace materials, processing methods, and components. This review mainly emphasises developments related to commercial passenger airplanes and light-duty cars. Improved and partly new materials for the reduction of airframe weight and joining by laser-beam welding and friction stir welding are ongoing areas of assessment. Chemical reactions, microstructure development, and residual stresses are frequently measured. Polymers and polymer matrix composites often require special experimental techniques. The thrust-to-weight ratio of aero-engines is increasing due to the improved design of components and the use of innovative materials. Investigations on superalloys, {gamma}-TiAl, and thermal barrier coatings are described in some detail. A discussion of the use of neutron and synchrotron diffraction in automotive applications covers the analysis of surface effects with respect to lubricants and wear, as well as the investigation of microstructure development, deformation, and fatigue behavior of materials, welds and components. Special steels, Al and Mg alloys are discussed and residual stresses in automotive components such as gears or crankshafts are described. Applications of characterization methods on membranes for polymeric membrane fuel cells and on nanocrystalline metal hydrides for hydrogen storage are shown. The degradation of railway tracks after long-term use is taken as an example for the application of synchrotron methods to transport systems beyond the commercial aircraft and light duty passenger car. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Material Challenges For Plasma Facing Components in Future Fusion Reactors

International Nuclear Information System (INIS)

Linke, J; Pintsuk, G.; Rödig, M.

2013-01-01

Increasing attention is directed towards thermonuclear fusion as a possible future energy source. Major advantages of this energy conversion technology are the almost inexhaustible resources and the option to produce energy without CO2-emissions. However, in the most advanced field of magnetic plasma confinement a number of technological challenges have to be met. In particular high-temperature resistant and plasma compatible materials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PFMs) and components (PFCs) in such fusion devices, i.e. the first wall (FW), the limiters and the divertor, are strongly affected by the plasma wall interaction processes and the applied intense thermal loads during plasma operation. On the one hand, these mechanisms have a strong influence on the plasma performance; on the other hand, they have major impact on the lifetime of the plasma facing armour. In present-day and next step devices the resulting thermal steady state heat loads to the first wall remain below 1 MWm-2; the limiters and the divertor are expected to be exposed to power densities being at least one order of magnitude above the FW-level, i.e. up to 20 MWm-2 for next step tokamaks such as ITER or DEMO. These requirements are responsible for high demands on the selection of qualified PFMs and heat sink materials as well as reliable fabrication processes for actively cooled plasma facing components. The technical solutions which are considered today are mainly based on the PFMs beryllium, carbon or tungsten joined to copper alloys or stainless steel heat sinks. In addition to the above mentioned quasi-stationary heat loads, short transient thermal pulses with deposited energy densities up to several tens of MJm-2 are a serious concern for next step tokamak devices. The most frequent events are so-called Edge Localized Modes (type I ELMs) and plasma disruptions

3D microcomputer tomograph for materials development and testing of components

International Nuclear Information System (INIS)

Riesemeier, H.; Goebbels, J.; Illerhaus, B.; Onel, Y.; Reimers, P.

1993-01-01

Examples prove the great capacity of 3D microcomputerized tomography in characterising new materials, particularly in the development stage. Cracks and delamination after deliberate damage are shown with good resolution. In showing complex structures, the possibility of picture reproduction of any sectional plane is of great use. 3D microcomputer tomography is therefore a futuristic process in materials research and for small of component series in production, eg: in the aircraft and space industry. (orig./DG) [de

PFMC-16. 16th international conference on plasma-facing materials and components for fusion applications. Abstracts

Energy Technology Data Exchange (ETDEWEB)

NONE

2017-07-01

The performances of fusion devices and of future fusion power plants strongly depend on the plasma-facing materials and components. Resistance to heat and particle loads, compatibility in plasma operations, thermo-mechanical properties, as well as the response to neutron irradiation are critical parameters which need to be understood and tailored from atomistic to component levels. The 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications addresses these issues.

Magnetic field effects on runaway electron energy deposition in plasma facing materials and components

International Nuclear Information System (INIS)

Niemer, K.A.; Gilligan, J.G.

1992-01-01

This paper reports magnetic field effects on runaway electron energy deposition in plasma facing materials and components is investigated using the Integrated TIGER Series. The Integrated TIGER Series is a set of time-independent coupled electron/photon Monte Carlo transport codes which perform photon and electron transport, with or without macroscopic electric and magnetic fields. A three-dimensional computational model of 100 MeV electrons incident on a graphite block was used to simulate runawayelectrons striking a plasma facing component at the edge of a tokamak. Results show that more energy from runaway electrons will be deposited in a material that is in the presence of a magnetic field than in a material that is in the presence of no field. For low angle incident runaway electrons in a strong magnetic field, the majority of the increased energy deposition is near the material surface with a higher energy density. Electrons which would have been reflected with no field, orbit the magnetic field lines and are redeposited in the material surface, resulting in a substantial increase in surface energy deposition. Based on previous studies, the higher energy deposition and energy density will result in higher temperatures which are expected to cause more damage to a plasma facing component

Application of Principal Component Analysis in Prompt Gamma Spectra for Material Sorting

Energy Technology Data Exchange (ETDEWEB)

Im, Hee Jung; Lee, Yun Hee; Song, Byoung Chul; Park, Yong Joon; Kim, Won Ho

2006-11-15

For the detection of illicit materials in a very short time by comparing unknown samples' gamma spectra to pre-programmed material signatures, we at first, selected a method to reduce the noise of the obtained gamma spectra. After a noise reduction, a pattern recognition technique was applied to discriminate the illicit materials from the innocuous materials in the noise reduced data. Principal component analysis was applied for a noise reduction and pattern recognition in prompt gamma spectra. A computer program for the detection of illicit materials based on PCA method was developed in our lab and can be applied to the PGNAA system for the baggage checking at all ports of entry at a very short time.

Research of application of new material to light water reactor components

International Nuclear Information System (INIS)

Mihara, Tanetoyo

1992-01-01

Advanced Nuclear Equipment Research Institute (ANERI) has been doing the research to apply the new material including metal, fine ceramics and high polymer which were developed and applied in other industries to components and parts of light water reactor for the purpose of Improvement of reliability of components, improvement of efficiency of periodic inspection, improvement of repair and reduction of radiation exposure of worker. This project started upon the sponsorship of Ministry of International Trade and Industry (MITI) by the schedule of FY1985-FY1993 (9 years) and effective results has been obtained. (author)

Fieldable Nuclear Material Identification System

International Nuclear Information System (INIS)

Radle, James E.; Archer, Daniel E.; Carter, Robert J.; Mullens, James Allen; Mihalczo, John T.; Britton, Charles L. Jr.; Lind, Randall F.; Wright, Michael C.

2010-01-01

The Fieldable Nuclear Material Identification System (FNMIS), funded by the NA-241 Office of Dismantlement and Transparency, provides information to determine the material attributes and identity of heavily shielded nuclear objects. This information will provide future treaty participants with verifiable information required by the treaty regime. The neutron interrogation technology uses a combination of information from induced fission neutron radiation and transmitted neutron imaging information to provide high confidence that the shielded item is consistent with the host's declaration. The combination of material identification information and the shape and configuration of the item are very difficult to spoof. When used at various points in the warhead dismantlement sequence, the information complimented by tags and seals can be used to track subassembly and piece part information as the disassembly occurs. The neutron transmission imaging has been developed during the last seven years and the signature analysis over the last several decades. The FNMIS is the culmination of the effort to put the technology in a usable configuration for potential treaty verification purposes.

Investigation on the micro injection molding process of an overmolded multi-material micro component

DEFF Research Database (Denmark)

Baruffi, Federico; Calaon, Matteo; Tosello, Guido

and difficult assembly steps, being the plastic molded directly on a metal substrate. In this scenario, an investigation on the fully automated micro overmolding manufacturing technology of a three-material micro component for acoustic applications has been carried out. Preliminary experiments allowed......Micro injection molding (μIM) is one of the few technologies capable of meeting the increasing demand of complex shaped micro plastic parts. This process, combined with the overmolding technique, allows a fast and cost-efficient production of multi-material micro components, saving numerous...

Handling Vagueness as an Intelligent Component of a Materials Information System.

Science.gov (United States)

Schudnagis, Monika; Womser-Hacker, Christa

1996-01-01

Discusses vagueness as a problem of materials information system development in the context of information retrieval within the paradigm of information science. Presents a prototype which combines an object-oriented graphical user interface with natural language feedback and correction functionality, as well as intelligent components for graphical…

Optimal base-stock levels in a serial two-echelon system with random yield and rework of orders

NARCIS (Netherlands)

Kiesmüller, G.P.; Kok, de A.G.

2007-01-01

In this paper we consider the manufacturing process of a finished product assembled out of one unit of an expensive subassembly and some other non-expensive parts. The subassembly itself is assembled out of several components delivered just in time from outside suppliers. After the production of the

Design considerations for multi component molecular-polymeric nonlinear optical materials

Energy Technology Data Exchange (ETDEWEB)

Singer, K.D. (Case Western Reserve Univ., Cleveland, OH (USA). Dept. of Physics); Kuzyk, M.G. (Washington State Univ., Pullman, WA (USA). Dept. of Physics); Fang, T.; Holland, W.R. (AT and T Bell Labs., Princeton, NJ (USA)); Cahill, P.A. (Sandia National Labs., Albuquerque, NM (USA))

1990-01-01

We review our work on multi component polymeric nonlinear optical materials. These materials consist of nonlinear optical molecules incorporated in a polymeric host. A cross-linked triazine polymer incorporating a dicyanovinyl terminated azo dye was found to be relatively stable at 85{degree} and posses an electro-optic coefficient of 11pm/V. We have also observed the zero dispersion condition in a new anomalous dispersion dye for phase matched second harmonic generation, and expect efficient conversion to the blue. A squarylium dye, ISQ, has been found to posses a large third order nonlinearity, and may display two-level behavior. 24 refs., 11 figs.

Advanced materials for critical components in industrial gas turbines

Energy Technology Data Exchange (ETDEWEB)

Gibbons, T.B. (Div. of Materials Metrology, National Physical Lab., Teddington (United Kingdom))

1992-06-01

Combined-cycle plant for power production has advantages in terms of capital costs and flexibility compared to large power plants either nuclear of fossil-fired, used for base load. In combined-cycle plant the overall efficiency is highly dependent on the performance of the gas turbine and turbine entry temperatures of > 1200deg C will be required to obtain attractive levels of efficiency. Bearing in mind the need for reliability and longterm performance from components such as turbine blades, the challenge to the materials enginer is formidable. In this paper some of the recent developments in Ni - Cr-base alloys are described and the potential for advanced materials such as ceramics and intermetallics is briefly considered. Development in coating technology to provide effective thermal barriers and good resistance to aggressive environments are discussed. (orig./MM).

Structural materials requirements for in-vessel components of fusion power plants

International Nuclear Information System (INIS)

Schaaf, B. van der

2000-01-01

The economic production of fusion energy is determined by principal choices such as using magnetic plasma confinement or generating inertial fusion energy. The first generation power plants will use deuterium and tritium mixtures as fuel, producing large amounts of highly energetic neutrons resulting in radiation damage in materials. In the far future the advanced fuels, 3 He or 11 B, determine power plant designs with less radiation damage than in the first generation. The first generation power plants design must anticipate radiation damage. Solid sacrificing armour or liquid layers could limit component replacements costs to economic levels. There is more than radiation damage resistance to determine the successful application of structural materials. High endurance against cyclic loading is a prominent requirement, both for magnetic and inertial fusion energy power plants. For high efficiency and compactness of the plant, elevated temperature behaviour should be attractive. Safety and environmental requirements demand that materials have low activation potential and little toxic effects under both normal and accident conditions. The long-term contenders for fusion power plant components near the plasma are materials in the range from innovative steels, such as reduced activation ferritic martensitic steels, to highly advanced ceramic composites based on silicon carbide, and chromium alloys. The steels follow an evolutionary path to basic plant efficiencies. The competition on the energy market in the middle of the next century might necessitate the riskier but more rewarding development of SiCSiC composites or chromium alloys

FOREWORD: 12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications 12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications

Science.gov (United States)

Kreter, Arkadi; Linke, Jochen; Rubel, Marek

2009-12-01

The 12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications (PFMC-12) was held in Forschungszentrum Jülich (FZJ) in Germany in May 2009. This symposium is the successor to the International Workshop on Carbon Materials for Fusion Applications series. Between 1985 and 2003, 10 'Carbon Workshops' were organized in Jülich, Stockholm and Hohenkammer. After this time, the scope of the symposium was redefined to reflect the new requirements of ITER and the ongoing evolution of the field. The workshop was first organized under its new name in 2006 in Greifswald, Germany. The main objective of this conference series is to provide a discussion forum for experts from research institutions and industry dealing with materials for plasma-facing components in present and future controlled fusion devices. The operation of ASDEX-Upgrade with tungsten-coated wall, the fast progress of the ITER-Like Wall Project at JET, the plans for the EAST tokamak to install tungsten, the start of ITER construction and a discussion about the wall material for DEMO all emphasize the importance of plasma-wall interactions and component behaviour, and give much momentum to the field. In this context, the properties and behaviour of beryllium, carbon and tungsten under plasma impact are research topics of foremost relevance and importance. Our community realizes both the enormous advantages and serious drawbacks of all the candidate materials. As a result, discussion is in progress as to whether to use carbon in ITER during the initial phase of operation or to abandon this element and use only metal components from the start. There is broad knowledge about carbon, both in terms of its excellent power-handling capabilities and the drawbacks related to chemical reactivity with fuel species and, as a consequence, about problems arising from fuel inventory and dust formation. We are learning continuously about beryllium and tungsten under fusion conditions, but our

In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

International Nuclear Information System (INIS)

Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

1983-01-01

Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

Fast Flux Test Facility (FFTF) feedback reactivity components

International Nuclear Information System (INIS)

Nguyen, D.H.

1988-04-01

The static tests conducted during Cycle 8A (1986) of the FFTF have allowed, for the first time, the experimental determination of each of the feedback reactivities caused by the following mechanisms: fuel axial expansion, control rod repositioning, core radial expansion, and subassembly bowing. A semiempirical equation was obtained to describe each of these feedback components that depended only on the relevant reactor temperature (bowing was presented in a tabular form). The Doppler and sodium density reactivities were calculated using existing mechanistic methods. Although they could also be fitted with closed-form equations depending only on temperatures, these equations are not needed in transient analyses using whole core safety computer codes, which use mechanistic methods. The static feedback reactivity model was extended to obtain a dynamic model via the concept of ''time constants.'' Besides being used for transient analyses in the FFTF, these feedback equations constitute a database for the validation and/or calibration of mechanistic feedback reactivity models. 2 refs., 6 tabs

New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

International Nuclear Information System (INIS)

Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

1996-01-01

The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

High-throughput quantum chemistry and virtual screening for OLED material components

Science.gov (United States)

Halls, Mathew D.; Giesen, David J.; Hughes, Thomas F.; Goldberg, Alexander; Cao, Yixiang

2013-09-01

Computational structure enumeration, analysis using an automated simulation workflow and filtering of large chemical structure libraries to identify lead systems, has become a central paradigm in drug discovery research. Transferring this paradigm to challenges in materials science is now possible due to advances in the speed of computational resources and the efficiency and stability of chemical simulation packages. State-of-the-art software tools that have been developed for drug discovery can be applied to efficiently explore the chemical design space to identify solutions for problems such as organic light-emitting diode material components. In this work, virtual screening for OLED materials based on intrinsic quantum mechanical properties is illustrated. Also, a new approach to more reliably identify candidate systems is introduced that is based on the chemical reaction energetics of defect pathways for OLED materials.

Selection of hardfacing material for components of the Indian Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Bhaduri, A.K.; Indira, R.; Albert, S.K.; Rao, B.P.S.; Jain, S.C.; Asokkumar, S.

2004-01-01

Nickel-base hardfacing alloys have been chosen to replace cobalt-base alloys as hardfacing material for components of the Indian Prototype Fast Breeder Reactor, for minimising the dose rate to personnel during maintenance and decommissioning, and to reduce the shielding thickness required for component handling. Induced activity, dose rate and shielding computations showed that replacing cobalt-base alloys with nickel-base alloys for hardfacing of components would result in a marked reduction in both the dose rate from the components and the thickness of lead handling flasks. Long-term ageing studies on the nickel-base hardface deposits on austenitic stainless steel showed that the hardface deposit would retain adequate hardness at the end of the components' design service-life of 40 years of exposure at 823 K

Plasma facing materials and components for future fusion devices - development, characterization and performance under fusion specific loading conditions

Energy Technology Data Exchange (ETDEWEB)

Linke, J. [Forschungszentrum Juelich (Germany). Inst. fuer Plasmaphysik

2006-04-15

The plasma exposed components in existing and future fusion devices are strongly affected by the plasma material interaction processes. These mechanisms have a strong influence on the plasma performance; in addition they have major impact on the lifetime of the plasma facing armour and the joining interface between the plasma facing material (PFM) and the heat sink. Besides physical and chemical sputtering processes, high heat quasi-stationary fluxes during normal and intense thermal transients are of serious concern for the engineers who develop reliable wall components. In addition, the material and component degradation due to intense fluxes of energetic neutrons is another critical issue in D-T-burning fusion devices which requires extensive RandD. This paper presents an overview on the materials development and joining, the testing of PFMs and components, and the analysis of the neutron irradiation induced degradation.

Plasma facing materials and components for future fusion devices - development, characterization and performance under fusion specific loading conditions

International Nuclear Information System (INIS)

Linke, J.

2006-01-01

The plasma exposed components in existing and future fusion devices are strongly affected by the plasma material interaction processes. These mechanisms have a strong influence on the plasma performance; in addition they have major impact on the lifetime of the plasma facing armour and the joining interface between the plasma facing material (PFM) and the heat sink. Besides physical and chemical sputtering processes, high heat quasi-stationary fluxes during normal and intense thermal transients are of serious concern for the engineers who develop reliable wall components. In addition, the material and component degradation due to intense fluxes of energetic neutrons is another critical issue in D-T-burning fusion devices which requires extensive RandD. This paper presents an overview on the materials development and joining, the testing of PFMs and components, and the analysis of the neutron irradiation induced degradation

Plasma facing materials and components for future fusion devices—development, characterization and performance under fusion specific loading conditions

Science.gov (United States)

Linke, J.

2006-04-01

The plasma exposed components in existing and future fusion devices are strongly affected by the plasma material interaction processes. These mechanisms have a strong influence on the plasma performance; in addition they have major impact on the lifetime of the plasma facing armour and the joining interface between the plasma facing material (PFM) and the heat sink. Besides physical and chemical sputtering processes, high heat quasi-stationary fluxes during normal and intense thermal transients are of serious concern for the engineers who develop reliable wall components. In addition, the material and component degradation due to intense fluxes of energetic neutrons is another critical issue in D-T-burning fusion devices which requires extensive R&D. This paper presents an overview on the materials development and joining, the testing of PFMs and components, and the analysis of the neutron irradiation induced degradation.

Influence of material choice on cost estimation of some key components of the Sulfur Iodine thermochemical process

International Nuclear Information System (INIS)

Gilardi, T.; Rodriguez, G.; Gomez, A.; Leybros, J.; Borgard, J.M.; Carles, P.; Anzieu, P.

2006-01-01

In the frame of the preliminary design of an sulfur/iodine thermochemical plant coupled with a 600 MWth Helium cooled High Temperature Reactor, CEA has pre-designed all the components of the I/S plant and has started to the cost estimation of all the key components with some industrial cost evaluation methods proposed by CHAUVEL or PETER and TIMMERHAUS. The purpose of the paper is to present the strong influence of material choice on final cost estimation of these key components by comparing price with standard material (steel) and the most appropriate material selected to support the strong corrosion involved by several chemical reactions of the I/S process. These results reinforce the fact that material selection must be done with the best accuracy and that it will be a key factor in the global economy of these plant investment. (authors)

Aspects for selection of materials and fabrication processes for nuclear component manufacturing

International Nuclear Information System (INIS)

Pernstich, K.

1980-01-01

For components of the Nuclear steam supply System of Light Water Reactors an extremely high safety standard is required. These requirements only can be met by adequate selection of materials and fabrication processes and their proper application in combination with strict quality assurance and control measurements. A general overview of the basic aspects to be considered in this connection is presented together with an indication of the present state of art for the main materials and fabrication processes. (author) [pt

Nuclear fuel string assembly

International Nuclear Information System (INIS)

Ip, A.K.; Koyanagi, K.; Tarasuk, W.R.

1976-01-01

A method of fabricating rodded fuels suitable for use in pressure tube type reactors and in pressure vessel type reactors is described. Fuel rods are secured as an inner and an outer sub-assembly, each rod attached between mounting rings secured to the rod ends. The two sub-assemblies are telescoped together and positioned by spaced thimbles located between them to provide precise positioning while permittng differential axial movement between the sub-assemblies. Such sub-assemblies are particularly suited for mounting as bundle strings. The method provides particular advantages in the assembly of annular-section fuel pins, which includes booster fuel containing enriched fuel material. (LL)

Costing the OMNIUM-G system 7500

Science.gov (United States)

Fortgang, H. R.

1980-01-01

A complete OMNIUM-G System 7500 was cost analyzed for annual production quantities ranging from 25 to 10,000 units per year. Parts and components were subjected to in-depth scrutiny to determine optimum manufacturing processes, coupled with make or buy decisions on materials and small parts. When production quantities increase both labor and material costs reduce substantially. A redesign of the system that was analyzed could result in lower costs when annual production runs approach 100,000 units/year. Material and labor costs for producing 25, 100, 25,000 and 100,00 units are given for 17 subassembly units.

The role of materials in the analysis of fast breeder reactor components

International Nuclear Information System (INIS)

Aubert, Michel; Petrequin, Pierre.

1982-09-01

The analysis of fast breeder reactor components involves the knowledge of certain properties of the materials used. The latter consist of the following: - a body of data required for calculations, including allowable stresses and fatigue strength, as well as the rules applicable to these data, - a number of qualitative requirements serving to guarantee that the quality of the material fully justifies the use of the previously established elements. This duality of concerns is illustrated by some recent examples which occured during the construction of the Super Phenix reactor [fr

Apparatus with moderating material for microwave heat treatment of manufactured components

Science.gov (United States)

Ripley, Edward B [Knoxville, TN

2011-05-10

An apparatus for heat treating manufactured components using microwave energy and microwave susceptor material. Heat treating medium such as eutectic salts may be employed. A fluidized bed introduces process gases which may include carburizing or nitriding gases The process may be operated in a batch mode or continuous process mode. A microwave heating probe may be used to restart a frozen eutectic salt bath.

Research on optimizing components of microfine high-performance composite cementitious materials

International Nuclear Information System (INIS)

Hu Shuguang; Guan Xuemao; Ding Qingjun

2002-01-01

The relationship between material components and mechanical properties was studied in terms of composite material principles and orthogonal experimental design. Moreover, the microstructure of microfine high-performance composite cementitious material (MHPCC) paste was investigated by means of scanning electron microscopy (SEM) methods. The results showed that the composite material consisting of blast furnace slag (BFS), gypsum (G 2 ) and expansive agent (EA) could obviously improve the strength of the cementitious material containing 40% fly ash (FA). Although microfine cement (MC) was merely 45% percent of the MHPCC, the compressive strength of MHPCC paste was higher than that of neat MC paste. BFS played an important role in MHPCC. The optimum-added quantity of BFS was 15%. The needle-shaped ettringite obtained from the EA reacting with Ca(OH) 2 forms a three-dimensional network structure, which not only improved the early strength of MHPCC paste but also increased its late strength. The reason was that the network structure, which was similar to a fiber-reinforced composite, was formed in the late period of hydration with the progress of hydration and the deposition of hydration products into the network structure

Magnetic fusion energy plasma interactive and high heat flux components. Volume III. Strategy for international collaborations in the areas of plasma materials interactions and high heat flux materials and components development

International Nuclear Information System (INIS)

Gauster, W.B.; Bauer, W.; Roberto, J.B.; Post, D.E.

1984-01-01

The purpose of this summary is to assess opportunities for such collaborations in the specific areas of Plasma Materials Interaction and High Heat Flux Materials and Components Development, and to aid in developing a strategy to take advantage of them. After some general discussion of international collaborations, we summarize key technical issues and the US programs to address them. Then follows a summary of present collaborations and potential opportunities in foreign laboratories

Preliminary cleaning tests on candidate materials for APS beamline and front end UHV components

International Nuclear Information System (INIS)

Nielsen, R.; Kuzay, T.M.

1992-01-01

Comparative cleaning tests have been done on four candidate materials for use in APS beamline and front-end vacuum components. These materials are 304 SS, 304L SS, OFHC copper, and Glidcop* (Cu-Al 2 O 3 )- Samples of each material were prepared and cleaned using two different methods. After cleaning, the sample surfaces were analyzed using ESCA (Electron Spectography for Chemical Analysis). Uncleaned samples were used as a reference. The cleaning methods and surface analysis results are further discussed

Design and optimization of components and processes for plasma sources in advanced material treatments

OpenAIRE

Rotundo, Fabio

2012-01-01

The research activities described in the present thesis have been oriented to the design and development of components and technological processes aimed at optimizing the performance of plasma sources in advanced in material treatments. Consumables components for high definition plasma arc cutting (PAC) torches were studied and developed. Experimental activities have in particular focussed on the modifications of the emissive insert with respect to the standard electrode configuration, whi...

A prototype knowledge-based system for material selection of ceramic matrix composites of automotive engine components

Energy Technology Data Exchange (ETDEWEB)

Sapuan, S.M.; Jacob, M.S.D.; Mustapha, F.; Ismail, N

2002-12-15

A prototype knowledge based system (KBS) for material selection of ceramic matrix composites (CMC) for engine components such as piston, connecting rod and piston ring is proposed in this paper. The main aim of this research work is to select the most suitable material for the automotive engine components. The selection criteria are based upon the pre-defined constraint value. The constraint values are mechanical, physical properties and manufacturing techniques. The constraint values are the safety values for the product design. The constraint values are selected from the product design specification. The product design specification values are selected from the past design calculation and some values are calculated by the help of past design data. The knowledge-based system consists of several modules such as knowledge acquisition module, inference module and user interface module. The domains of the knowledge-based system are defined as objects and linked together by hierarchical graph. The system is capable of selecting the most suitable materials and ranks the materials with respect to their properties. The design engineers can choose the required materials related to the materials property.

Behaviour of alkali halides as materials for optical components of high power lasers

International Nuclear Information System (INIS)

Apostol, D.I.; Mihailescu, N.I.; Ghiordanescu, V.; Nistor, C.L.; Nistor, V.S.; Teodorescu, V.; Voda, M.

1978-01-01

The physical phenomena taking place in alkali halides when a CO 2 laser radiation is passing through have been reviewed. A special emphasis has been put on the specific qualities which such materials should have for being used as components for high power lasers. (author)

Suitability of Co as an alloy material for components of the primary circuit of HTR reactors

International Nuclear Information System (INIS)

Iniotakis, N.

1977-02-01

For high temperature reactors it is of interest if Co-alloys could be used for the different components of the primary cooling circuit. It has been investigated in detail to what amount the Co-60 created by neutron activation of Co-59 contained in the material of the components could possibly contribute to the contamination of the primary cooling circuit of the reactor. The result of these investigations is compared with the contamination of the cooling circuit by fission and activation products like Co-137, Cs-134, Ag-11om etc. For pebble bed reactors with an OTTO-type fuel management it could be shown that there is no limitation for the use of cobalt in alloys for materials of the components in the primary cooling circuit. The only boundary condition is that the local Thermal Flux at the position of the components should be less than phisub(th) 7 n/cm 2 . sec. (orig.) [de

Materials and Components for Low Temperature Solid Oxide Fuel Cells – an Overview

Directory of Open Access Journals (Sweden)

D. Radhika

2013-06-01

Full Text Available This article summarizes the recent advancements made in the area of materials and components for low temperature solid oxide fuel cells (LT-SOFCs. LT-SOFC is a new trend in SOFCtechnology since high temperature SOFC puts very high demands on the materials and too expensive to match marketability. The current status of the electrolyte and electrode materials used in SOFCs, their specific features and the need for utilizing them for LT-SOFC are presented precisely in this review article. The section on electrolytes gives an overview of zirconia, lanthanum gallate and ceria based materials. Also, this review article explains the application of different anode, cathode and interconnect materials used for SOFC systems. SOFC can result in better performance with the application of liquid fuels such methanol and ethanol. As a whole, this review article discusses the novel materials suitable for operation of SOFC systems especially for low temperature operation.

Evaluation of material integrity on electricity generator water steam cycles component (Main Steam Pipe)

International Nuclear Information System (INIS)

Sudardjo; Histori; Triyadi, Ari

1998-01-01

The evaluation of material integrity on electricity generator component has been done. That component was main steam pipe of Unit II Suralaya Coal Fired Power Plant. evaluation was done by replication technique. The damage was found are two porosity's, from two point samples of six points sample population. Based on cavity evaluation in steels, which proposed by Neubauer and Wedel that porosity's still at class A damage. For class A damage, its means no remedial action would be required until next major scheduled maintenance outage. That porosity's was grouped on isolated cavities and not need ti repair that main steam pipe component less than three year after replication test

R and D program for French sodium fast reactor: On the description and detection of sodium boiling phenomena during sub-assembly blockages

International Nuclear Information System (INIS)

Vanderhaegen, M.; Paumel, K.; Seiler, J. M.; Tourin, A.; Jeannot, J. P.; Rodriguez, G.

2011-01-01

In support of the French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor program, which aims to demonstrate the industrial applicability of sodium fast reactors with an increased level of safety demonstration and availability compared to the past French sodium fast reactors, emphasis is placed on reactor instrumentation. It is in this framework that CEA studies continuous core monitoring to detect as early as possible the onset of sodium boiling. Such a detection system is of particular interest due to the rapid progress and the consequences of a Total Instantaneous Blockage (TIB) at a subassembly inlet, where sodium boiling intervenes in an early phase. In this paper, the authors describe all the particularities which intervene during the different boiling stages and explore possibilities for their detection. (authors)

Contour integral computations for multi-component material systems subjected to creep

International Nuclear Information System (INIS)

Chen, J.-J.; Tu, S.-T.; Xuan, F.-Z.; Wang, Z.-D.

2006-01-01

In the present paper the crack behavior of multi-component material systems is investigated under extensive creep condition. The validation of the creep fracture parameters C* and C(t) is firstly examined at the microscale level. It is found that the C* value is no longer path-independent when mismatch inclusions are embedded into the matrix. To characterize the crack fields in inhomogeneous material the integral value defined at the crack tip as C tip * is introduced to reflect the influence of the inclusion. The interaction effects between microcrack and inclusion are systematically calculated with respect to different mismatch factors, various inclusion locations and inclusion numbers. The analysis results show that the C tip * value is not only influenced by the inclusion properties but also depends on the microstructure near the crack tip

Structural materials for ITER in-vessel component design

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

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

Supply of equipment and components, materials of dedicated commercial grade, other services

International Nuclear Information System (INIS)

Perdices, D.; Perez Medina, E.

2014-01-01

The following article describes the course of action of Tecnatom as Procurement Agent and Manufacturing Manager for the supply materials, equipment and components for the Spanish Nuclear Power Plants. We have devoted a special mention to the supply of dedicated commercial grade items (CGI), bringing together the services of Manufacturing Manager, Engineering service and testing facilities, simplifying the control of the supply chain with total warranty. (Author)

Experimental investigation of surface determination process on multi-material components for dimensional computed tomography

DEFF Research Database (Denmark)

Borges de Oliveira, Fabrício; Stolfi, Alessandro; Bartscher, Markus

2016-01-01

The possibility of measuring multi-material components, while assessing inner and outer features simultaneously makes X-ray computed tomography (CT) the latest evolution in the field of coordinate measurement systems (CMSs). However, the difficulty in selecting suitable scanning parameters and su...

Fine 3D neutronic characterization of a gas-cooled fast reactor based on plate-type sub-assemblies

International Nuclear Information System (INIS)

Bosq, J. C.; Peneliau, Y.; Rimpault, G.; Vanier, M.

2006-01-01

CEA neutronic studies have allowed the definition of a first 2400 MWth reference gas-cooled fast reactor core using plate-type sub-assemblies, for which the main neutronic characteristics were calculated by the so-called ERANOS 'design calculation scheme' relying on several method approximations. The last stage has consisted in a new refine characterization, using the reference calculation scheme, in order to confirm the impact of the approximations of the design route. A first core lay-out taking into account control rods was proposed and the reactivity penalty due to the control rod introduction in this hexagonal core lay-out was quantified. A new adjusted core was defined with an increase of the plutonium content. This leads to a significant decrease of the breeding gain which needs to be recovered in future design evolutions in order to achieve the self breeding goal. Finally, the safety criteria associated to the control rods were calculated with a first estimation of the uncertainties. All these criteria are respected, even if the safety analysis of GFR concepts and the determination of these uncertainties should be further studied and improved. (authors)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Calculations of the possible consequences of molten fuel sodium interactions in subassembly and whole core geometries

International Nuclear Information System (INIS)

Coddington, P.; Fishlock, T.P.; Jakeman, D.

1976-01-01

The possible consequences of molten fuel sodium interactions are calculated using various modelling assumptions and key parameters. And the significance of the choice of assumptions and parameters are discussed. As for subassembly geometry, the results of one-dimensional code EXPEL are compared with the solutions of the one-dimensional Lagrangian equations of a compressible fluid (TOPAL was used). The adequacy of acoustic approximation used in EXPEL is discussed here. The effects of heat transfer time constant on the behaviour of peak pressure are also analyzed by parametric surveys. Other items investigated are the length and position of the interacting zone, the existence of a non-condensable gas volume, and the vapour condensation on cold clad. As for whole core geometry, a simple dynamical model of arc expanding spherical interacting zone immersed in a semi-infinite sea of cold liquid was used (SHORE code). Within the interacting zone a simple heat transfer model (including a heat transfer time and a fragmentation time) was adopted. Vapour blanketing was considered in a number of ways. Representative results of the calculations are given in a table. Containment studies were also performed for ''ducted'' design and ''open pool'' design. The development of new codes in the U.K. for these analysis are also briefly described. (Aoki, K.)

Towards a more consolidated approach to material data management in life assessment of power plant components

Energy Technology Data Exchange (ETDEWEB)

Jovanovic, A; Maile, K [MPA Stuttgart (Germany)

1999-12-31

The presentation discusses the necessity of having a more consolidated (unified, possibly `European`) framework for all (not only pure experimental) material data needed for optimized life management and assessment of high-temperature and other components in power and process plants. After setting the main requirements for such a system, a description of efforts done in this direction at MPA Stuttgart in the area of high-temperature components in power plants is given. Furthermore, a reference to other relevant efforts elsewhere is made and an example of practical application of the proposed solution described (optimized material selection and life assessment of high-temperature piping). (orig.) 10 refs.

Towards a more consolidated approach to material data management in life assessment of power plant components

Energy Technology Data Exchange (ETDEWEB)

Jovanovic, A.; Maile, K. [MPA Stuttgart (Germany)

1998-12-31

The presentation discusses the necessity of having a more consolidated (unified, possibly `European`) framework for all (not only pure experimental) material data needed for optimized life management and assessment of high-temperature and other components in power and process plants. After setting the main requirements for such a system, a description of efforts done in this direction at MPA Stuttgart in the area of high-temperature components in power plants is given. Furthermore, a reference to other relevant efforts elsewhere is made and an example of practical application of the proposed solution described (optimized material selection and life assessment of high-temperature piping). (orig.) 10 refs.

Effect of operating conditions and environment on properties of materials of PWR type nuclear power plant components

International Nuclear Information System (INIS)

Vacek, M.

1987-01-01

Operating reliability and service life of PWR type nuclear power plants are discussed with respect to the material properties of the plant components. The effects of the operating environment on the material properties and the methods of their determination are characterized. Discussed are core materials, such as fuel, its cladding and regulating rod materials, and the materials of pipes, steam generators and condensers. The advances in the production of pressure vessel materials and their degradation during operation are treated in great detail. (Z.M.)

Disseny d’un component bi-material pel sector de l’automòbil

OpenAIRE

Igual Bueno, Jordi

2017-01-01

El projecte fi de grau que es desenvolupa en aquest document mostra el treball que implica el disseny d’un component bi-material del sector de l’automoció. I a més, mostra les diferents etapes de construcció de les màquines-eines que fan viable la producció en sèrie de la peça desenvolupada. La peça que es treballa en aquest projecte és un Front End que és un component que està ubicat en la part oculta anterior del vehicle. Inicialment s’analitzen els fitxers rebuts per part del client i e...

Flow induced vibration studies on PFBR control plug components

Energy Technology Data Exchange (ETDEWEB)

Prakash, V., E-mail: prakash@igcar.gov.in [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamilnadu (India); Kumar, P. Anup; Anandaraj, M.; Thirumalai, M.; Anandbabu, C.; Rajan, K.K. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamilnadu (India)

2012-09-15

Highlights: Black-Right-Pointing-Pointer Flow induced vibration studies on Prototype Fast Breeder Reactor control plug model carried out. Black-Right-Pointing-Pointer Velocity similitude was followed for the study. Black-Right-Pointing-Pointer Frequencies and amplitude of vibrations of various control plug components measured. Black-Right-Pointing-Pointer Overall values of vibration are well within permissible limits. - Abstract: The construction of Prototype Fast Breeder Reactor (PFBR), a 500 MWe liquid sodium cooled reactor, is in progress at Kalpakkam in India. Control plug (CP) is located right above the core subassemblies in the hot pool. Control plug is an important component as many of the critical reactor parameters are sensed and controlled by the components housed in the control plug assembly. In PFBR primary circuit, components are basically thin walled, slender shells with diameter to thickness ratio ranging from 100 to 650. These components are prone to flow induced vibrations. The existence of free liquid (sodium) surfaces, which is the source of sloshing phenomenon and the operation of primary sodium pump in the primary pool are other potential sources of vibration of reactor components. Control plug is a hollow cylindrical shell structure and provides passages and support for 12 absorber rod drive mechanisms (ARDM) which consists of 9 control and safety rods and 3 diverse safety rods, 210 thermo wells to measure the sodium temperature at the exit of various fuel subassemblies, three failed fuel localization modules (FFLM) and acoustic detectors. It consists of a core cover plate (CCP), which forms the bottom end, two intermediate supports plate, i.e. lower stay plate (LSP) and upper stay plate (USP) and an outer shell. The CCP is located at a distance of 1.3 m from the core top. With such a gap, there will be long free hanging length of the thermocouple sleeves, Delayed neutron detector (DND) sampling tubes and ARDM shroud tubes and hence they are

Recycle and reuse of materials and components from waste streams of nuclear fuel cycle facilities

International Nuclear Information System (INIS)

2000-01-01

All nuclear fuel cycle processes utilize a wide range of equipment and materials to produce the final products they are designed for. However, as at any other industrial facility, during operation of the nuclear fuel cycle facilities, apart from the main products some byproducts, spent materials and waste are generated. A lot of these materials, byproducts or some components of waste have a potential value and may be recycled within the original process or reused outside either directly or after appropriate treatment. The issue of recycle and reuse of valuable material is important for all industries including the nuclear fuel cycle. The level of different materials involvement and opportunities for their recycle and reuse in nuclear industry are different at different stages of nuclear fuel cycle activity, generally increasing from the front end to the back end processes and decommissioning. Minimization of waste arisings and the practice of recycle and reuse can improve process economics and can minimize the potential environmental impact. Recognizing the importance of this subject, the International Atomic Energy Agency initiated the preparation of this report aiming to review and summarize the information on the existing recycling and reuse practice for both radioactive and non-radioactive components of waste streams at nuclear fuel cycle facilities. This report analyses the existing options, approaches and developments in recycle and reuse in nuclear industry

Stress analysis of glass-ceramic insulator and molybdenum cylinders in vacuum tube subassembly

International Nuclear Information System (INIS)

Spears, R.K.

1980-01-01

This study determined the state of stress between molybdenum cylinders and a glass-ceramic insulator of a vacuum tube during cooling when the glass-ceramic coefficient of expansion differed from molybdenum by +-2 x 10 -7 / 0 C. A thermoelastic stress analysis was performed on the vacuum tube subassembly using the finite element method. Two cases, which examined the effect of cooling over a 700 0 C range, were considered. In Case One, the expansion coefficient of the glass-ceramic was 2 x 10 -7 / 0 C less than that of molybdenum while for Case Two, it was 2 x 10 -7 / 0 C greater. For Case One, it was found that the tangential stresses in the insulator were entirely compressive but the maximum principal stresses in the r-z plane were mainly tensile. For Case Two, the tangential stresses were tensile in the insulator as were most of the maximum principal stresses in the r-z plane except for stress in the upper regions of the insulator. The magnitude of the stress at the maximum principal stress location appears to be substantially lower than what has been observed in practice (i.e., cracking of this design had never been a major problem, but it has been observed that if the coefficient of expansion of the glass-ceramic was 2 x 10 -7 / 0 C lower than molybdenum, cracking usually resulted). This analysis showed that the expansion coefficient of the glass-ceramic could be varied quite liberally from molybdenum before the ultimate strength (13,000 lb/in. 2 ) of the glass-ceramic was exceeded

Performance of tungsten-based materials and components under ITER and DEMO relevant steady-state thermal loads

Energy Technology Data Exchange (ETDEWEB)

Ritz, Guillaume Henri

2011-07-01

In nuclear fusion devices the surfaces directly facing the plasma are irradiated with high energy fluxes. The most intense loads are deposited on the divertor located at the bottom of the plasma chamber, which has to withstand continuous heat loads with a power density of several MW . m{sup -2} as well as transient events. These are much shorter (in the millisecond and sub-millisecond regime) but deposit a higher power densities of a few GW . m{sup -2}. The search for materials that can survive to those severe loading conditions led to the choice of tungsten which possesses advantageous attributes such as a high melting point, high thermal conductivity, low thermal expansion and an acceptable activation rate. These properties made it an attractive and promising candidate as armor material for divertors of future fusion devices such as ITER and DEMO. For the DEMO divertor, conceptual studies on helium-cooled tungsten plasma-facing components were performed. The concept was realized and tested under DEMO specific cyclic thermal loads. The examination of the plasma-facing components by microstructural analyses before and after thermal loading enabled to determine the mechanisms for components failure. Among others, it clearly showed the impact of the tungsten grade and the thermal stress induced crack formation on the performance of the armor material and in general of the plasma-facing component under high heat loads. A tungsten qualification program was launched to study the behaviour of various tungsten grades, in particular the crack formation, under fusion relevant steady-state thermal loads. In total, seven commercially available materials from two industrial suppliers were investigated. As the material's thermal response is strongly related to its microstructure, this program comprised different material geometries and manufacturing technologies. It also included the utilization of an actively cooled specimen holder which has been designed to perform

Performance of tungsten-based materials and components under ITER and DEMO relevant steady-state thermal loads

International Nuclear Information System (INIS)

Ritz, Guillaume Henri

2011-01-01

In nuclear fusion devices the surfaces directly facing the plasma are irradiated with high energy fluxes. The most intense loads are deposited on the divertor located at the bottom of the plasma chamber, which has to withstand continuous heat loads with a power density of several MW . m -2 as well as transient events. These are much shorter (in the millisecond and sub-millisecond regime) but deposit a higher power densities of a few GW . m -2 . The search for materials that can survive to those severe loading conditions led to the choice of tungsten which possesses advantageous attributes such as a high melting point, high thermal conductivity, low thermal expansion and an acceptable activation rate. These properties made it an attractive and promising candidate as armor material for divertors of future fusion devices such as ITER and DEMO. For the DEMO divertor, conceptual studies on helium-cooled tungsten plasma-facing components were performed. The concept was realized and tested under DEMO specific cyclic thermal loads. The examination of the plasma-facing components by microstructural analyses before and after thermal loading enabled to determine the mechanisms for components failure. Among others, it clearly showed the impact of the tungsten grade and the thermal stress induced crack formation on the performance of the armor material and in general of the plasma-facing component under high heat loads. A tungsten qualification program was launched to study the behaviour of various tungsten grades, in particular the crack formation, under fusion relevant steady-state thermal loads. In total, seven commercially available materials from two industrial suppliers were investigated. As the material's thermal response is strongly related to its microstructure, this program comprised different material geometries and manufacturing technologies. It also included the utilization of an actively cooled specimen holder which has been designed to perform sophisticated

Standardization Efforts for Mechanical Testing and Design of Advanced Ceramic Materials and Components

Science.gov (United States)

Salem, Jonathan A.; Jenkins, Michael G.

2003-01-01

Advanced aerospace systems occasionally require the use of very brittle materials such as sapphire and ultra-high temperature ceramics. Although great progress has been made in the development of methods and standards for machining, testing and design of component from these materials, additional development and dissemination of standard practices is needed. ASTM Committee C28 on Advanced Ceramics and ISO TC 206 have taken a lead role in the standardization of testing for ceramics, and recent efforts and needs in standards development by Committee C28 on Advanced Ceramics will be summarized. In some cases, the engineers, etc. involved are unaware of the latest developments, and traditional approaches applicable to other material systems are applied. Two examples of flight hardware failures that might have been prevented via education and standardization will be presented.

Ti-doped isotropic graphite: A promising armour material for plasma-facing components

Science.gov (United States)

García-Rosales, C.; López-Galilea, I.; Ordás, N.; Adelhelm, C.; Balden, M.; Pintsuk, G.; Grattarola, M.; Gualco, C.

2009-04-01

Finely dispersed Ti-doped isotropic graphites with 4 at.% Ti have been manufactured using synthetic mesophase pitch 'AR' as raw material. These new materials show a thermal conductivity at room temperature of ˜200 W/mK and flexural strength close to 100 MPa. Measurement of the total erosion yield by deuterium bombardment at ion energies and sample temperatures for which pure carbon shows maximum values, resulted in a reduction of at least a factor of 4, mainly due to dopant enrichment at the surface caused by preferential erosion of carbon. In addition, ITER relevant thermal shock loads were applied with an energetic electron beam at the JUDITH facility. The results demonstrated a significantly improved performance of Ti-doped graphite compared to pure graphite. Finally, Ti-doped graphite was successfully brazed to a CuCrZr block using a Mo interlayer. These results let assume that Ti-doped graphite can be a promising armour material for divertor plasma-facing components.

Ti-doped isotropic graphite: A promising armour material for plasma-facing components

Energy Technology Data Exchange (ETDEWEB)

Garcia-Rosales, C. [CEIT and Tecnun (University of Navarra), Paseo de Manuel Lardizabal, 15, E-20018 San Sebastian (Spain)], E-mail: cgrosales@ceit.es; Lopez-Galilea, I.; Ordas, N. [CEIT and Tecnun (University of Navarra), Paseo de Manuel Lardizabal, 15, E-20018 San Sebastian (Spain); Adelhelm, C.; Balden, M. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, D-85748 Garching (Germany); Pintsuk, G. [Forschungszentrum Juelich GmbH, EURATOM Association, D-52425 Juelich (Germany); Grattarola, M.; Gualco, C. [Ansaldo Ricerche S.p.A., I-16152 Genoa (Italy)

2009-04-30

Finely dispersed Ti-doped isotropic graphites with 4 at.% Ti have been manufactured using synthetic mesophase pitch 'AR' as raw material. These new materials show a thermal conductivity at room temperature of {approx}200 W/mK and flexural strength close to 100 MPa. Measurement of the total erosion yield by deuterium bombardment at ion energies and sample temperatures for which pure carbon shows maximum values, resulted in a reduction of at least a factor of 4, mainly due to dopant enrichment at the surface caused by preferential erosion of carbon. In addition, ITER relevant thermal shock loads were applied with an energetic electron beam at the JUDITH facility. The results demonstrated a significantly improved performance of Ti-doped graphite compared to pure graphite. Finally, Ti-doped graphite was successfully brazed to a CuCrZr block using a Mo interlayer. These results let assume that Ti-doped graphite can be a promising armour material for divertor plasma-facing components.

Ti-doped isotropic graphite: A promising armour material for plasma-facing components

International Nuclear Information System (INIS)

Garcia-Rosales, C.; Lopez-Galilea, I.; Ordas, N.; Adelhelm, C.; Balden, M.; Pintsuk, G.; Grattarola, M.; Gualco, C.

2009-01-01

Finely dispersed Ti-doped isotropic graphites with 4 at.% Ti have been manufactured using synthetic mesophase pitch 'AR' as raw material. These new materials show a thermal conductivity at room temperature of ∼200 W/mK and flexural strength close to 100 MPa. Measurement of the total erosion yield by deuterium bombardment at ion energies and sample temperatures for which pure carbon shows maximum values, resulted in a reduction of at least a factor of 4, mainly due to dopant enrichment at the surface caused by preferential erosion of carbon. In addition, ITER relevant thermal shock loads were applied with an energetic electron beam at the JUDITH facility. The results demonstrated a significantly improved performance of Ti-doped graphite compared to pure graphite. Finally, Ti-doped graphite was successfully brazed to a CuCrZr block using a Mo interlayer. These results let assume that Ti-doped graphite can be a promising armour material for divertor plasma-facing components.

A review of the United Kingdom fast reactor programme

International Nuclear Information System (INIS)

Bramman, J.I.; Hickey, H.B.; Wheeler, R.C.; Gregory, C.V.

1989-01-01

The total electricity generating capacity in the UK is approximately 54 GW. Total electricity generation in 1988 was 288 TW hours, of which just over 20% was nuclear. In Scotland the percentage of electricity generated by nuclear stations was 49% of the total, and will exceed 60% in 1989. The privatization of the Electricity Supply Industry (ESI) in the UK (mentioned in last year's report) is proceeding on schedule. Considerable efforts are being made to ensure that the maximum benefits will be obtained from operating the PFR during the next five years. The main thrust of the UKAEA's programme continues to be towards the requirements of the EFR. Reload 16 included the biennial maintenance and statutory inspection period. It was extended from its original 60 days by the need to carry out modifications aimed at improving the reliability of the protection systems designed to safeguard the components of the secondary circuit, including the IHXs, in the event of a sodium-water reaction in a steam generator unit, and by the need to inspect and repair the vessels of the steam generator units. Good progress was made with the fuel development programme. The leading experimental cluster of PE16-clad 6.6 mm diameter pins is continuing irradiation above 21% burnup and 150 dpa (NRT). The lead subassembly with 5.8 mm pins clad in PE16 has exceeded 17.6% burnup, 130 dpa (NRT). The leading subassembly with pins of the same type to have undergone complete PIE contained fuel at 16% burnup and PE16 clad at 116 dpa; these pins were found to be in very good condition. Radial blanket subassemblies have exceeded 2% burnup without failure. In 1988/89 there was one reprocessing campaign in the PFR Reprocessing Plant lasting from November 1988 to February 1989. Feed material included irradiated fuel from 12 subassemblies irradiated in the PFR, some unirradiated subassemblies and loose pins and residues; in all containing 1.3t of Heavy Metal (HM) containing 242 kg plutonium. The cumulative

Materials for the plasma-facing components of steady state stellarators

International Nuclear Information System (INIS)

Bolt, H.; Boscary, J.; Greuner, H.; Grigull, P.; Maier, H.; Streibl, B.

2005-01-01

, stellarators as well as tokamaks, the situation is likely to be different. The long operation times of several years between refurbishment shut downs and the low neutron irradiation resistance will most likely prevent the use of low-Z materials. Tungsten as a main candidate high-Z material is presently being intensely investigated. The properties of tungsten coatings and of massive tungsten as well as the related component technology will be discussed in the presentation. (author)

Ceramic-Based 4D Components: Additive Manufacturing (AM) of Ceramic-Based Functionally Graded Materials (FGM) by Thermoplastic 3D Printing (T3DP).

Science.gov (United States)

Scheithauer, Uwe; Weingarten, Steven; Johne, Robert; Schwarzer, Eric; Abel, Johannes; Richter, Hans-Jürgen; Moritz, Tassilo; Michaelis, Alexander

2017-11-28

In our study, we investigated the additive manufacturing (AM) of ceramic-based functionally graded materials (FGM) by the direct AM technology thermoplastic 3D printing (T3DP). Zirconia components with varying microstructures were additively manufactured by using thermoplastic suspensions with different contents of pore-forming agents (PFA), which were co-sintered defect-free. Different materials were investigated concerning their suitability as PFA for the T3DP process. Diverse zirconia-based suspensions were prepared and used for the AM of single- and multi-material test components. All of the samples were sintered defect-free, and in the end, we could realize a brick wall-like component consisting of dense (<1% porosity) and porous (approx. 5% porosity) zirconia areas to combine different properties in one component. T3DP opens the door to the AM of further ceramic-based 4D components, such as multi-color, multi-material, or especially, multi-functional components.

Design of the ITER Tokamak Assembly Tools

International Nuclear Information System (INIS)

Park, Hyunki; Her, Namil; Kim, Byungchul; Im, Kihak; Jung, Kijung; Lee, Jaehyuk; Im, Kisuk

2006-01-01

ITER (International Thermonuclear Experimental Reactor) Procurement allocation among the seven Parties, EU, JA, CN, IN , KO, RF and US had been decided in Dec. 2005. ITER Tokamak assembly tools is one of the nine components allocated to Korea for the construction of the ITER. Assembly tools except measurement and common tools are supplied to assemble the ITER Tokamak and classified into 9 groups according to components to be assembled. Among the 9 groups of assembly tools, large-sized Sector Sub-assembly Tools and Sector Assembly Tools are used at the first stage of ITER Tokamak construction and need to be designed faster than seven other assembly tools. ITER IT (International Team) proposed Korea to accomplish ITA (ITER Transitional Arrangements) Task on detailed design, manufacturing feasibility and contract specification of specific, large sized tools such as Upending Tool, Lifting Tool, Sector Sub-assembly Tool and Sector Assembly Tool in Oct. 2004. Based on the concept design by ITER IT, Korea carried out ITA Task on detailed design of large-sized and specific Sector Sub-assembly and Sector Assembly Tools until Mar. 2006. The Sector Sub-assembly Tools mainly consist of the Upending, Lifting, Vacuum Vessel Support and Bracing, and Sector Sub-assembly Tool, among which the design of three tools are herein. The Sector Assembly Tools mainly consist of the Toroidal Field (TF) Gravity Support Assembly, Sector In-pit Assembly, TF Coil Assembly, Vacuum Vessel (VV) Welding and Vacuum Vessel Thermal Shield (TS) Assembly Tool, among which the design of Sector In-pit Assembly Tool is described herein

On the use of elastic-plastic material characteristics for linear-elastic component assessments

International Nuclear Information System (INIS)

Kussmaul, K.; Silcher, H.; Eisele, U.

1995-01-01

In this paper the procedure of safety assessment of components by fracture mechanics analysis as recommended in TECDOC 717 is applied to two standard specimens of ductile cast iron. It is shown that the use of a pseudo-elastic K IJ -value in linear elastic safety analysis may lead to non-conservative results, when elastic-plastic material behaviour can be expected. (author)

Identification and Assessment of Material Models for Age-Related Degradation of Structures and Passive Components in Nuclear Power Plants

International Nuclear Information System (INIS)

Choi, In Kil; Kim, Min Kyu; Hofmayer, Charles; Braverman, Joseph; Nie, Jinsuo

2009-03-01

This report describes the research effort performed by BNL for the Year 2 scope of work. This research focused on methods that could be used to represent the long-term behavior of materials used at NPPs. To achieve this BNL reviewed time-dependent models which can approximate the degradation effects of the key materials used in the construction of structures and passive components determined to be of interest in the Year 1 effort. The intent was to review the degradation models that would cover the most common time-dependent changes in material properties for concrete and steel components

Material selection for an aerospace component

OpenAIRE

Jönsson, Gustav

2015-01-01

In the world of today there is a drive for lighter and more effective products for various reasons e.g. reduced environmental impact, higher payload, fuel efficiency etc. There is also an expanding development of new materials for a large number of different applications. This makes it more and more difficult for engineers to make good material selections. This has led to the development of a large amount of material selection methods that require more or less effort to select material. An ef...

The application of prepared porous carbon materials: Effect of different components on the heavy metal adsorption.

Science.gov (United States)

Song, Min; Wei, Yuexing; Yu, Lei; Tang, Xinhong

2016-06-01

In this study, five typical municipal solid waste (MSW) components (tyres, cardboard, polyvinyl chloride (PVC), acrylic textile, toilet paper) were used as raw materials to prepare four kinds of MSW-based carbon materials (paperboard-based carbon materials (AC1); the tyres and paperboard-based carbon materials (AC2); the tyres, paperboard and PVC-based carbon materials (AC3); the tyres, paperboard, toilet paper, PVC and acrylic textile-based carbon materials (AC4)) by the KOH activation method. The characteristic results illustrate that the prepared carbon adsorbents exhibited a large pore volume, high surface area and sufficient oxygen functional groups. Furthermore, the application of AC1, AC2, AC3, AC4 on different heavy metal (Cu(2+), Zn(2+), Pb(2+), Cr(3+)) removals was explored to investigate their adsorption properties. The effects of reaction time, pH, temperature and adsorbent dosage on the adsorption capability of heavy metals were investigated. Comparisons of heavy metal adsorption on carbon of different components were carried out. Among the four samples, AC1 exhibits the highest adsorption capacity for Cu(2+); the highest adsorption capacities of Pb(2+) and Zn(2+) are obtained for AC2; that of Cr(3+) are obtained for AC4. In addition, the carbon materials exhibit better adsorption capability of Cu(2+) and Pb(2+) than the other two kind of metal ions (Zn(2+) and Cr(3+)). © The Author(s) 2016.

Repair process and a repaired component

Energy Technology Data Exchange (ETDEWEB)

Roberts, III, Herbert Chidsey; Simpson, Stanley F.

2018-02-20

Matrix composite component repair processes are disclosed. The matrix composite repair process includes applying a repair material to a matrix composite component, securing the repair material to the matrix composite component with an external securing mechanism and curing the repair material to bond the repair material to the matrix composite component during the securing by the external securing mechanism. The matrix composite component is selected from the group consisting of a ceramic matrix composite, a polymer matrix composite, and a metal matrix composite. In another embodiment, the repair process includes applying a partially-cured repair material to a matrix composite component, and curing the repair material to bond the repair material to the matrix composite component, an external securing mechanism securing the repair material throughout a curing period, In another embodiment, the external securing mechanism is consumed or decomposed during the repair process.

Ageing studies on materials, components and process instruments used in nuclear power plants

International Nuclear Information System (INIS)

Bora, J.S.

1997-04-01

This report is a compilation of test results of thermal and radiation ageing tests carried out in the laboratory over a period of 25 years on diverse engineering materials, components and instruments used in nuclear power plants. Test items covered are different types of electrical cables, elastomers, surface coatings, electrical and electronics components and process instruments. Effects of thermal and radiation ageing on performance parameters are shown in tabular forms. Apart from finding the characteristics, capabilities and limitations of test items, ageing research has helped in pin-pointing sub-standard and critical parts and necessary corrective action has been taken. This report is expected to be quite useful to the manufacturers users and researchers for reference and guidance. (author)

STRUCTURAL DESIGN CRITERIA FOR TARGET/BLANKET SYSTEM COMPONENT MATERIALS FOR THE ACCELERATOR PRODUCTION OF TRITIUM PROJECT

Energy Technology Data Exchange (ETDEWEB)

W. JOHNSON; R. RYDER; P. RITTENHOUSE

2001-01-01

The design of target/blanket system components for the Accelerator Production of Tritium (APT) plant is dependent on the development of materials properties data specified by the designer. These data are needed to verify that component designs are adequate. The adequacy of the data will be related to safety, performance, and economic considerations, and to other requirements that may be deemed necessary by customers and regulatory bodies. The data required may already be in existence, as in the open technical literature, or may need to be generated, as is often the case for the design of new systems operating under relatively unique conditions. The designers' starting point for design data needs is generally some form of design criteria used in conjunction with a specified set of loading conditions and associated performance requirements. Most criteria are aimed at verifying the structural adequacy of the component, and often take the form of national or international standards such as the ASME Boiler and Pressure Vessel Code (ASME B and PV Code) or the French Nuclear Structural Requirements (RCC-MR). Whether or not there are specific design data needs associated with the use of these design criteria will largely depend on the uniqueness of the conditions of operation of the component. A component designed in accordance with the ASME B and PV Code, where no unusual environmental conditions exist, will utilize well-documented, statistically-evaluated developed in conjunction with the Code, and will not be likely to have any design data needs. On the other hand, a component to be designed to operate under unique APT conditions, is likely to have significant design data needs. Such a component is also likely to require special design criteria for verification of its structural adequacy, specifically accounting for changes in materials properties which may occur during exposure in the service environment. In such a situation it is common for the design criteria

Cylinder components properties, applications, materials

CERN Document Server

2016-01-01

Owing to the ever-increasing requirements to be met by gasoline and diesel engines in terms of CO2 reduction, emission behavior, weight, and service life, a comprehensive understanding of combustion engine components is essential today. It is no longer possible for professionals in automotive engineering to manage without the corresponding expertise, whether they work in the field of design, development, testing, or maintenance. This technical book provides in-depth answers to questions about design, production, and machining of cylinder components. In this second edition, every section has been revised and expanded to include the latest developments in the combustion engine. Content Piston rings Piston pins and piston pin circlips Bearings Connecting rods Crankcase and cylinder liners Target audience Engineers in the field of engine development and maintenanceLecturers and students in the areas of mechanical engineering, engine technology, and vehicle constructionAnyone interested in technology Publisher MAH...

Process for surface treatment of zirconium-containing cladding materials for fuel element or other components for nuclear reactors

International Nuclear Information System (INIS)

Videm, K.G.; Lunde, L.R.; Kooyman, H.H.

1975-01-01

A process for the surface treatment of zirconium-base cladding materials for fuel elements or other components for nuclear reactors is described. The treatment includes pickling the cladding material in a fluoride-containing bath, and then applying a protective coating through oxidation to the pickled cladding material. The fluoride-containing contaminants which remain on the surface of the cladding material during pickling are removed or rendered harmless by anodic oxidation

Calculations of the Possible Consequences of Molten Fuel Sodium Interactions in Subassembly and Whole Core Geometries

International Nuclear Information System (INIS)

Coddington, P.; Fishlock, T.P.; Jakeman, D.

1976-01-01

In making assessments of fast reactor safety a number of accident sequences can be postulated in which molten fuel contacts sodium in a number of possible modes. In the absence of an understanding of the way in which reactor materials interact for these contact modes it is necessary to make assessments over a range of plausible conditions and assumptions. This enables those areas where an interaction might cause a new stage in the escalation of the accident to be identified and at the same time to establish what characteristics of the interaction may be important. Whether in real situations interaction of molten reactor materials can have such characteristics can then be considered from both a theoretical and experimental viewpoint. It is suggested that although high efficiency vapour explosions involving large amounts of fuel in which there is rapid and coherent fragmentation are a main source of concern in many accident sequences, interactions with other characteristics may also be important. Two areas which have been identified are: (i) the interactions of low efficiency which need only involve small fractions of the fuel or possibly could include molten clad but which can accelerate sodium and fuel sufficiently to give rise to large reactivity changes. The recent incident at a steel plant in the U.K. in which 100 tons of molten steel was ejected to a height of 10 m from a torpedo ladle when water accidentally poured into it is a particularly striking illustration of such movement; and (ii) interactions giving rise to a much slower and less coherent heat transfer which may require some degree of fragmentation but not the extensive fragmentation by the specific mechanisms associated with vapour explosions but which nevertheless on the reactor scale could lead to high slug impacts on the containment. Accident codes are being constructed in the U.K. to investigate a series of hypothetical incidents. Modules are required for these codes which enable the consequences

Comparison of measured and calculated composition of irradiated EBR-II blanket assemblies

International Nuclear Information System (INIS)

Grimm, K. N.

1998-01-01

In anticipation of processing irradiated EBR-II depleted uranium blanket subassemblies in the Fuel Conditioning Facility (FCF) at ANL-West, it has been possible to obtain a limited set of destructive chemical analyses of samples from a single EBR-II blanket subassembly. Comparison of calculated values with these measurements is being used to validate a depletion methodology based on a limited number of generic models of EBR-II to simulate the irradiation history of these subassemblies. Initial comparisons indicate these methods are adequate to meet the operations and material control and accountancy (MC and A) requirements for the FCF, but also indicate several shortcomings which may be corrected or improved

Improved Materials for Use as Components in Kraft Black Liquor Recovery Boilers; TOPICAL

International Nuclear Information System (INIS)

Keiser, J.R.

2001-01-01

This Cooperative Research and Development Agreement (CRADA) was undertaken to evaluate current and improved materials and materials processing conditions for use as components in kraft black liquor recovery boilers and other unit processes. The main areas addressed were: (1) Improved Black Liquor Nozzles, (2) Weld Overlay of Composite Floor Tubes, and (3) Materials for Lime Kilns. Iron aluminide was evaluated as an alternate material for the nozzles used to inject an aqueous solution known as black liquor into recovery boilers as well for the uncooled lining in the ports used for the nozzles. Although iron aluminide is known to have much better sulfidation resistance in gases than low alloy and stainless steels, it did not perform adequately in the environment where it came into contact with molten carbonate, sulfide and sulfate salts. Weld overlaying carbon steel tubes with a layer of stainless weld metal was a proposed method of extending the life of recovery boiler floor tubes that have experienced considerable fireside corrosion. After exposure under service conditions, sections of weld overlaid floor tubes were removed from a boiler floor and examined metallographically. Examination results indicated satisfactory performance of the tubes. Refractory-lined lime kilns are a critical component of the recovery process in kraft pulp mills, and the integrity of the lining is essential to the successful operation of the kiln. A modeling study was performed to determine the cause of, and possible solutions for, the repeated loss of the refractory lining from the cooled end of a particular kiln. The evaluation showed that the temperature, the brick shape and the coefficient of friction between the bricks were the most important parameters influencing the behavior of the refractory lining

Rapsodie first core manufacture. 1. part: processing plant

International Nuclear Information System (INIS)

Masselot, Y.; Bataller, S.; Ganivet, M.; Guillet, H.; Robillard, A.; Stosskopf, F.

1968-01-01

This report is the first in a series of three describing the processes, results and peculiar technical problems related to the manufacture of the first core of the fast reactor Rapsodie. A detailed study of manufacturing processes(pellets, pins, fissile sub-assemblies), the associated testings (raw materials, processed pellets and pins, sub-assemblies before delivery), manufacturing facilities and improvements for a second campaign are described. (author) [fr

Component design and testing for a miniaturised autonomous sensor based on a nanowire materials platform

NARCIS (Netherlands)

Rajesh Ramaneti; Francois Krummenacher; Fritz Falk; Naser Khosropour; Björn Eisenhawer; Cees van Rijn; Giorgos Fagas; Ran Yu; Adrian M. Ionescu; Ing. Erik Puik; Montserrat Fernández-Bolaños Badia; Nikolay Petkov; Hien Duy Tong; Rik Lafeber; John C De Mello; Olan Lotty; Adrian M. Nightingale; Yordan M. Georgiev; Elizabeth Buitrago; Frank van der Bent; Michael Nolan; Justin D. Holmes; Annett Gawlik; Maher Kayal; Guobin Jia

2014-01-01

From Springer description: "We present the design considerations of an autonomous wireless sensor and discuss the fabrication and testing of the various components including the energy harvester, the active sensing devices and the power management and sensor interface circuits. A common materials

Division of Development and Technology Plasma/Materials Interaction and High Heat Flux Materials and Components Task Groups: Report on the joint meeting, July 9, 1986

International Nuclear Information System (INIS)

Watson, R.D.

1986-09-01

This paper contains a collection of viewgraphs from a joint meeting of the Division of Development and Technology Plasma/Materials Interaction and High Heat Flux Materials and Components Task Groups. A list of contributing topics is: PPPL update, ATF update, Los Alamos RFP program update, status of DIII-D, PMI graphite studies at ORNL, PMI studies for low atomic number materials, high heat flux materials issues, high heat flux testing program, particle confinement in tokamaks, helium self pumping, self-regenerating coatings technical planning activity and international collaboration update

Materials, manufacture and testing of pressurized components of high-power steam power plants

International Nuclear Information System (INIS)

Blind, D.; Foehl, J.; Issler, L.; Schellhammer, W.; Sturm, D.; Kussmaul, K.; Heinrich, D.; Meyer, H.J.; Prestel, W.

1981-01-01

This is the first German review of materials, production and testing of pressure components of high-capacity steam power plants. The authors have been working in this field for years; their special subject has been the availability and reliability of pressure vessels, in particular in nuclear engineering. Fundamentals are presented as well as the findings obtained at the state Materials Testing Institute in Stuttgart. The material is presented in a well-structured classification; the most recent international findings, especially of the USA, are presented. This is possible due to the close cooperation between the Stuttgart institute and a number of US research institutes. The new subject of fracture mechanics is treated in some detail; its fundamentals are discussed from the American point of view while German considerations - in particular of the Reactor Safety Commission - are taken into account in the field of applications. (orig.) [de

The Jefferson Lab Quality Assurance Program for the SNS Superconducting Linac Construction Project

International Nuclear Information System (INIS)

Joseph Ozelis

2003-01-01

As part of a multi-laboratory collaboration, Jefferson Lab is currently engaged in the fabrication, assembly, and testing of 23 cryomodules for the superconducting linac portion of the Spallation Neutron Source (SNS) being built at Oak Ridge National Laboratory. As with any large accelerator construction project, it is vitally important that these components be built in a cost effective and timely manner, and that they meet the stringent performance requirements dictated by the project specifications. A comprehensive Quality Assurance (QA) program designed to help accomplish these goals has been implemented as an inherent component of JLab's SNS construction effort. This QA program encompasses the traditional spectrum of component performance, from incoming parts inspection, raw materials testing, through to sub-assembly and finished article performance evaluation

Design, construction and operating experience of demonstration LMFBRs. The application of core and fuel performance experience in British reactors to commercial fast reactor design

International Nuclear Information System (INIS)

Bagley, K.Q.

1978-01-01

The Prototype Fast Reactor (PFR) sub-assembly design is described with particular emphasis on the choice of factors that are important in determining satisfactory performance. Reasons for the adoption of specific clad and fuel design details are given in their historical context, and irradiation experience - mostly from the Dounreay Fast Reactor (DFR) - in support of the choices is described. The implications of factors that are now better understood than when the PFR fuel was designed, notably neutron-induced void swelling and irradiation creep, are then considered. It is shown that the 'free-standing' core design used in PFR, in which the sub-assembly is unsupported above the level of the lower axial breeder, relies on the availability of low-swelling, preferably irradiation-creep-resistant alloys as sub-assembly structural materials in order to achieve the prescribed burn-up target. The advantages of a 'restrained core', which makes use of irradiation creep to redress the effects of material swelling, are noted briefly, and the application of this concept to the Commercial Demonstration Fast Reactor (CDFR) core design is described. Probable future trends in pin and sub-assembly design are reviewed and the scope of associated irradiation testing programmes defined. Arrangements for monitoring and evaluating fuel performance, both in reactor and post-irradiation, are outlined and the provisions for endorsement of CDFR pin, sub-assembly and core design details in PFR are indicated. (author)

Design and synthesis of single-source molecular precursors to homogeneous multi-component oxide materials

Science.gov (United States)

Fujdala, Kyle Lee

This dissertation describes the syntheses of single-source molecular precursors to multi-component oxide materials. These molecules possess a core metal or element with various combinations of -OSi(O tBu)3, -O2P(OtBu) 2, and -OB[OSi(OtBu)3] 2 ligands. Such molecules decompose under mild thermolytic conditions (models for oxide-supported metal species and multi-component oxides. Significantly, the first complexes to contain three or more heteroelements suitable for use in the TMP method have been synthesized. Compounds for use as single-source molecular precursors have been synthesized containing Al, B, Cr, Hf, Mo, V, W, and Zr, and their thermal transformations have been examined. Heterogeneous catalytic reactions have been examined for selected materials. Also, cothermolyses of molecular precursors and additional molecules (i.e., metal alkoxides) have been utilized to provide materials with several components for potential use as catalysts or catalyst supports. Reactions of one and two equivs of HOSi(OtBu) 3 with Cr(OtBu)4 afforded the first Cr(IV) alkoxysiloxy complexes (tBuO) 3CrOSi(OtBu)3 and ( tBuO)2Cr[OSi(OtBu) 3]2, respectively. The high-yielding, convenient synthesis of (tBuO)3CrOSi(O tBu)3 make this complex a useful single-source molecular precursor, via the TMP method, to Cr/Si/O materials. The thermal transformations of (tBuO)3CrOSi(O tBu)3 and (tBuO) 2Cr[OSi(OtBu)3]2 to chromia-silica materials occurr at low temperatures (≤180°C), to give isobutene as the major carbon-containing product. The material generated from the solid-state conversion of (tBuO) 3CrOSi(OtBu)3 (CrOS ss) has an unexpectedly high surface area of 315 m2 g-1 that is slightly reduced to 275 m2 g-1 after calcination at 500°C in O2. The xerogel obtained by the thermolysis of an n-octane solution of (tBuO)3CrOSi(O tBu)3 (CrOSixg) has a surface area of 315 m2 g-1 that is reduced to 205 m2 g-1 upon calcination at 500°C. Powder X-ray diffraction (PXRD) analysis revealed that Cr2O 3 is

Manipulator for plasma-assisted machining of components made of materials with low machinability

International Nuclear Information System (INIS)

Lyaoshchukov, M.M.; Agadzhanyan, R.A.

1984-01-01

The All-Union Scientific-Research and Technological Institute of Pump Engineering developed, and the ''Uralgidromash'' Production Association has adopted, a manipulator with remote control for the plasma-assisted machining (PAM) of components made of materials with low machinability. The manipulator is distinguished by its universal design and can be used for machining both external and internal surfaces of the bodies of revolution and also end faces and various curvilinear surfaces

FOREWORD: 13th International Workshop on Plasma-Facing Materials and Components for Fusion Applications/1st International Conference on Fusion Energy Materials Science 13th International Workshop on Plasma-Facing Materials and Components for Fusion Applications/1st International Conference on Fusion Energy Materials Science

Science.gov (United States)

Jacob, Wolfgang; Linsmeier, Christian; Rubel, Marek

2011-12-01

The 13th International Workshop on Plasma-Facing Materials and Components (PFMC-13) jointly organized with the 1st International Conference on Fusion Energy Materials Science (FEMaS-1) was held in Rosenheim (Germany) on 9-13 May 2011. PFMC-13 is a successor of the International Workshop on Carbon Materials for Fusion Applications series. Between 1985 and 2003 ten 'Carbon Workshops' were organized in Jülich, Stockholm and Hohenkammer. Then it was time for a change and redefinition of the scope of the symposium to reflect the new requirements of ITER and the ongoing evolution in the field. Under the new name (PFMC-11), the workshop was first organized in 2006 in Greifswald, Germany and PFMC-12 took place in Jülich in 2009. Initially starting in 1985 with about 40 participants as a 1.5 day workshop, the event has continuously grown to about 220 participants at PFMC-12. Due to the joint organization with FEMaS-1, PFMC-13 set a new record with more than 280 participants. The European project Fusion Energy Materials Science, FEMaS, coordinated by the Max-Planck-Institut für Plasmaphysik (IPP), organizes and stimulates cooperative research activities which involve large-scale research facilities as well as other top-level materials characterization laboratories. Five different fields are addressed: benchmarking experiments for radiation damage modelling, the application of micro-mechanical characterization methods, synchrotron and neutron radiation-based techniques and advanced nanoscopic analysis based on transmission electron microscopy. All these fields need to be exploited further by the fusion materials community for timely materials solutions for a DEMO reactor. In order to integrate these materials research fields, FEMaS acted as a co-organizer for the 2011 workshop and successfully introduced a number of participants from research labs and universities into the PFMC community. Plasma-facing materials experience particularly hostile conditions as they are

A Conceptual Design and Structural Analysis for ITER Mid-plane Brace Tools

Energy Technology Data Exchange (ETDEWEB)

Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Jae Hyuk; Kim, Kyung kyu [SFA Engineering Corp., Changwon (Korea, Republic of); Im, Kihak; Robert, Shaw [ITER Organization, St Paul lez Durance Cedex (France)

2010-10-15

The ITER, International Thermonuclear Experimental Reactor, Tokamak machine is mainly composed of 9 vacuum vessel (VV)/toroidal field coils (TFCs)/vacuum vessel thermal shields (VVTS) 40 .deg. sectors. Each VV/TFCs/VVTS 40 .deg. sector is made up of one 40 .deg. VV, two 20 .deg. TFCs and associated VVTS segments. The ITER Tokamak assembly tools are purpose-built tools to assemble the ITER Tokamak machine which includes the cryostat and the components contained therein. Based on the design description document prepared by the IO (ITER international organization), Korea has carried out the conceptual design of assembly tools with IO cooperation. The 40 .deg. sector assemblies attached mid-plane brace tools sub-assembled at assembly hall are transferred to Tokamak hall using the lifting tool operated by Tokamak main cranes. The sector sub-assembly tools are composed of the upending tool, the sector sub-assembly tool, the sector lifting tool and the vacuum vessel support and mid-plane brace tools. The mid-plane brace tool is assembled to inner surface of VV and TFCs in phase of sector sub-assembly after completion of all sector components. VV, TFC and VVTS are separated fully before completion of 9 sectors at Tokamak in-pit. In this paper the mid-plane brace tools is introduced about function, structure and status of research and development are also described

Automated packaging platform for low-cost high-performance optical components manufacturing

Science.gov (United States)

Ku, Robert T.

2004-05-01

Delivering high performance integrated optical components at low cost is critical to the continuing recovery and growth of the optical communications industry. In today's market, network equipment vendors need to provide their customers with new solutions that reduce operating expenses and enable new revenue generating IP services. They must depend on the availability of highly integrated optical modules exhibiting high performance, small package size, low power consumption, and most importantly, low cost. The cost of typical optical system hardware is dominated by linecards that are in turn cost-dominated by transmitters and receivers or transceivers and transponders. Cost effective packaging of optical components in these small size modules is becoming the biggest challenge to be addressed. For many traditional component suppliers in our industry, the combination of small size, high performance, and low cost appears to be in conflict and not feasible with conventional product design concepts and labor intensive manual assembly and test. With the advent of photonic integration, there are a variety of materials, optics, substrates, active/passive devices, and mechanical/RF piece parts to manage in manufacturing to achieve high performance at low cost. The use of automation has been demonstrated to surpass manual operation in cost (even with very low labor cost) as well as product uniformity and quality. In this paper, we will discuss the value of using an automated packaging platform.for the assembly and test of high performance active components, such as 2.5Gb/s and 10 Gb/s sources and receivers. Low cost, high performance manufacturing can best be achieved by leveraging a flexible packaging platform to address a multitude of laser and detector devices, integration of electronics and handle various package bodies and fiber configurations. This paper describes the operation and results of working robotic assemblers in the manufacture of a Laser Optical Subassembly

Hydraulic experiments on the failed fuel location module of prototype fast breeder reactor

International Nuclear Information System (INIS)

Rajesh, K.; Kumar, S.; Padmakumar, G.; Prakash, V.; Vijayashree, R.; Rajan Babu, V.; Govinda Rajan, S.; Vaidyanathan, G.; Prabhaker, R.

2003-01-01

The design of Prototype Fast Breeder Reactor (PFBR) is based on sound design concepts with emphasis on intrinsic safety. The uncertainties involved in the design of various components, which are difficult to assess theoretically, are experimentally verified before design is validated. In PFBR core, the coolant (liquid sodium) enters the bottom of the fuel subassembly, passes over the fuel pins picking up the fission heat and issues in to a hot pool. If there is any breach in the fuel pins, the fission products come in direct contact with the coolant. This is undesirable and it is necessary to locate the subassembly with the failed fuel pin and to isolate it. A component called Failed Fuel Location Module (FFLM) is employed for locating the failed SA by monitoring the coolant samples coming out of each Subassembly. The coolant sample from each Subassembly is drawn by FFLM using an EM pump through sampling tube and selector valve and is monitored for the presence of delayed neutrons which is an indication of failure of the Subassembly. The pressure drop across the selector valve determines the rating of the EM Pump. The dilution of the coolant sample across the selector valve determines the effectiveness of monitoring for contamination. It is not possible to predict pressure drop across the selector valve and dilution of the coolant sample theoretically. These two parameters are determined using a hydraulic experiment on the FFLM. The experiment was carried out in conditions that simulate the reactor conditions following appropriate similarity laws. The paper discusses the details of the model, techniques of experiments and the results from the studies

Estrogen Receptor Binding Affinity of Food Contact Material Components Estimated by QSAR.

Science.gov (United States)

Sosnovcová, Jitka; Rucki, Marián; Bendová, Hana

2016-09-01

The presented work characterized components of food contact materials (FCM) with potential to bind to estrogen receptor (ER) and cause adverse effects in the human organism. The QSAR Toolbox, software application designed to identify and fill toxicological data gaps for chemical hazard assessment, was used. Estrogen receptors are much less of a lock-and-key interaction than highly specific ones. The ER is nonspecific enough to permit binding with a diverse array of chemical structures. There are three primary ER binding subpockets, each with different requirements for hydrogen bonding. More than 900 compounds approved as of FCM components were evaluated for their potential to bind on ER. All evaluated chemicals were subcategorized to five groups with respect to the binding potential to ER: very strong, strong, moderate, weak binder, and no binder to ER. In total 46 compounds were characterized as potential disturbers of estrogen receptor. Among the group of selected chemicals, compounds with high and even very high affinity to the ER binding subpockets were found. These compounds may act as gene activators and cause adverse effects in the organism, particularly during pregnancy and breast-feeding. It should be considered to carry out further in vitro or in vivo tests to confirm their potential to disturb the regulation of physiological processes in humans by abnormal ER signaling and subsequently remove these chemicals from the list of approved food contact materials. Copyright© by the National Institute of Public Health, Prague 2016

Scintillation crystal mounting apparatus

International Nuclear Information System (INIS)

Engdahl, L.W.; Deans, A.J.

1982-01-01

An improved detector head for a gamma camera is disclosed. The detector head includes a housing and a detector assembly mounted within the housing. Components of the detector assembly include a crystal sub-assembly, a phototube array, and a light pipe between the phototube array and crystal sub-assembly. The invention provides a unique structure for maintaining the phototubes in optical relationship with the light pipe and preventing the application of forces that would cause the camera's crystal to crack

Leading research on supermetals. Part 3. Small component material; Supermetal no sendo kenkyu. 3. Kogata buzai

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-03-01

The control technology of nanostructure and amorphous structure for small components was studied by literature and patent survey. On nanostructured heat-resistant high-strength materials, a particle dispersion technology by MA method, up-grading technology of steel materials by controlled rolling and oxide metallurgy, and grain refinement technology by phase transformation were extracted. On nanostructured light-weight high-strength malarial, Al base alloy is promising from the viewpoint of resource and energy saving. On amorphous high environment resistant materials, the study on the corrosion resistance of Mo-Ta, Fe-Cr and Ni-Fe alloys is promising. On amorphous magnetic materials, the structure system with an amorphous formability higher than that of previous soft magnetic materials was found. R & D subjects for creating advanced amorphous and nanostructured bulk materials were extracted. Development of supporting technologies for material creation, formation and evaluation is necessary as well as the material design/control technology. 569 refs., 85 figs., 11 tabs.

Multicriteria Decision Analysis in Improving Quality of Design in Femoral Component of Knee Prostheses: Influence of Interface Geometry and Material

Directory of Open Access Journals (Sweden)

Ali Jahan

2015-01-01

Full Text Available Knee prostheses as medical products require careful application of quality and design tool to ensure the best performance. Therefore, quality function deployment (QFD was proposed as a quality tool to systematically integrate consumer’s expectation to perceived needs by medical and design team and to explicitly address the translation of customer needs into engineering characteristics. In this study, full factorial design of experiment (DOE method was accompanied by finite element analysis (FEA to evaluate the effect of inner contours of femoral component on mechanical stability of the implant and biomechanical stresses within the implant components and adjacent bone areas with preservation of the outer contours for standard Co-Cr alloy and a promising functionally graded material (FGM. The ANOVA revealed that the inner shape of femoral component influenced the performance measures in which the angle between the distal and anterior cuts and the angle between the distal and posterior cuts were greatly influential. In the final ranking of alternatives, using multicriteria decision analysis (MCDA, the designs with FGM was ranked first over the Co-Cr femoral component, but the original design with Co-Cr material was not the best choice femoral component, among the top ranked design with the same material.

The Feed Materials Program of the Manhattan Project: A Foundational Component of the Nuclear Weapons Complex

Science.gov (United States)

Reed, B. Cameron

2014-12-01

The feed materials program of the Manhattan Project was responsible for procuring uranium-bearing ores and materials and processing them into forms suitable for use as source materials for the Project's uranium-enrichment factories and plutonium-producing reactors. This aspect of the Manhattan Project has tended to be overlooked in comparison with the Project's more dramatic accomplishments, but was absolutely vital to the success of those endeavors: without appropriate raw materials and the means to process them, nuclear weapons and much of the subsequent cold war would never have come to pass. Drawing from information available in Manhattan Engineer District Documents, this paper examines the sources and processing of uranium-bearing materials used in making the first nuclear weapons and how the feed materials program became a central foundational component of the postwar nuclear weapons complex.

Multi-Axial Simulation Table (MAST)

Data.gov (United States)

Federal Laboratory Consortium — The MAST delivers an extensive array of testing applications providing rapid, flexible and reliable analysis for ground vehicle components and subassemblies. Using...

The impact of duct-to-duct interaction on the hex duct dilation

International Nuclear Information System (INIS)

Lee, M.J.; Chang, L.K.; Lahm, C.E.; Porter, D.L.

1992-01-01

Dilation of the hex duct is an important factor in the operational lifetime of fuel subassemblies in liquid metal fast reactors. It is caused primarily by the irradiation-enhanced creep and void swelling of the hex duct material. Excessive dilation may jeopardize subassembly removal from the core or cause a subassembly storage problem where the grid size of the storage basket is limited. Dilation of the hex duct in Experimental Breeder Reactor II (EBR-II) limits useful lifetime because of these storage basket limitations. It is, therefore, important to understand the hex duct dilation behavior to guide the design and in-core management of fuel subassemblies in a way that excessive duct deformation can be avoided. To investigate the dilation phenomena, finite-element models of the hex duct have been developed. The inelastic analyses were performed using the structural analysis code, ANSYS. Both Type 316 and D9 austenitic stainless steel ducts are considered. The calculated dilations are in good agreement with profilometry measurements made after irradiation. The analysis indicates that subassembly interaction is an important parameter in addition to neutron fluence and temperature in determining hex duct dilation. 5 refs

XHM-1 alloy as a promising structural material for water-cooled fusion reactor components

International Nuclear Information System (INIS)

Solonin, M.I.; Alekseev, A.B.; Kazennov, Yu.I.; Khramtsov, V.F.; Kondrat'ev, V.P.; Krasina, T.A.; Rechitsky, V.N.; Stepankov, V.N.; Votinov, S.N.

1996-01-01

Experience gained in utilizing austenitic stainless steel components in water-cooled power reactors indicates that the main cause of their failure is the steel's propensity for corrosion cracking. In search of a material immune to this type of corrosion, different types of austenitic steels and chromium-nickel alloys were investigated and tested at VNIINM. This paper presents the results of studying physical and mechanical properties, irradiation and corrosion resistance in a water coolant at <350 C of the alloy XHM-1 as compared with austenitic stainless steels 00Cr16Ni15Mo3Nb, 00Cr20Ni25Nb and alloy 00Cr20Ni40Mo5Nb. Analysis of the results shows that, as distinct from the stainless steels studied, the XHM-1 alloy is completely immune to corrosion cracking (CC). Not a single induced damage was encountered within 50 to 350 C in water containing different amounts of chlorides and oxygen under tensile stresses up to the yield strength of the material. One more distinctive feature of the alloy compared to steels is that no change in the strength or total elongation is encountered in the alloy specimens irradiated to 32 dpa at 350 C. The XHM-1 alloy has adequate fabricability and high weldability characteristics. As far as its properties are concerned, the XHM-1 alloy is very promising as a material for water-cooled fusion reactor components. (orig.)

Feasibility study on measuring axial and transverse stress/strain components in composite materials using Bragg sensors

Science.gov (United States)

Luyckx, G.; Degrieck, J.; De Waele, W.; Van Paepegem, W.; Van Roosbroeck, J.; Chah, K.; Vlekken, J.; McKenzie, I.; Obst, A.

2017-11-01

A fibre optic sensor design is proposed for simultaneously measuring the 3D stress (or strain) components and temperature inside thermo hardened composite materials. The sensor is based on two fibre Bragg gratings written in polarisation maintaining fibre. Based on calculations of the condition number, it will be shown that reasonable accuracies are to be expected. First tests on the bare sensors and on the sensors embedded in composite material, which confirm the expected behaviour, will be presented.

Elaboration of functionally graded materials for plasma facing components of the thermonuclear machines

International Nuclear Information System (INIS)

Autissier, Emmanuel

2014-01-01

The objective of this study was to develop a Functionally Graded Material (FGM) W/Cu to replace the compliance layer (Cu-OFHC) in the plasma facing components of thermonuclear fusion reactor like ITER. The peculiarity of this work is to elaborate these materials without exceeding the melting temperature of copper in order to control its microstructure. The co-sintering is the most attractive solution to achieve this goal. The first phase of this study has been to decrease the sintering temperature of the tungsten to achieve this co-sintering. The elaboration of a Functionally Graded Materials being delicate, thermomechanical calculations were performed in order to determine the number and chemical composition in order to increase the lifespan of Plasma Facing Components. Spark Plasma Sintering conditions were optimized in order to achieve maximum density of W x Cu 1-x composites. The effect of copper content and density of the W x Cu 1-x composites on thermal and mechanical properties was investigated. The SPS conditions were applied for W/CuCrZr assemblies with a compliance layer composed of several interlayers. The importance of time for the integrity of assemblies thereof has been highlighted. The study of the dwell time during W/CuCrZr assembly leads to identify a parameter to characterize the integrity of the interface regardless of the composition and the nature of the layer of compliance. Moreover, the phenomena associated with the formation of the interface assembly have been identified. The interface W/W x Cu 1-x is formed by the extrusion of the copper layer of the W x Cu 1-x inside the tungsten porosities. The W y Cu 1-y /CuCrZr interface is formed by copper migration of CuCrZr layer inside the W y Cu 1-y layer. Finally optimization assembly conditions showed that the mechanical stresses due to the densification of the Functionally Graded Materials can be limited by sintering the FGM before the assembly. (author)

Mid-frequency Band Dynamics of Large Space Structures

Science.gov (United States)

Coppolino, Robert N.; Adams, Douglas S.

2004-01-01

High and low intensity dynamic environments experienced by a spacecraft during launch and on-orbit operations, respectively, induce structural loads and motions, which are difficult to reliably predict. Structural dynamics in low- and mid-frequency bands are sensitive to component interface uncertainty and non-linearity as evidenced in laboratory testing and flight operations. Analytical tools for prediction of linear system response are not necessarily adequate for reliable prediction of mid-frequency band dynamics and analysis of measured laboratory and flight data. A new MATLAB toolbox, designed to address the key challenges of mid-frequency band dynamics, is introduced in this paper. Finite-element models of major subassemblies are defined following rational frequency-wavelength guidelines. For computational efficiency, these subassemblies are described as linear, component mode models. The complete structural system model is composed of component mode subassemblies and linear or non-linear joint descriptions. Computation and display of structural dynamic responses are accomplished employing well-established, stable numerical methods, modern signal processing procedures and descriptive graphical tools. Parametric sensitivity and Monte-Carlo based system identification tools are used to reconcile models with experimental data and investigate the effects of uncertainties. Models and dynamic responses are exported for employment in applications, such as detailed structural integrity and mechanical-optical-control performance analyses.

Other components

International Nuclear Information System (INIS)

Anon.

1993-01-01

This chapter includes descriptions of electronic and mechanical components which do not merit a chapter to themselves. Other hardware requires mention because of particularly high tolerance or intolerance of exposure to radiation. A more systematic analysis of radiation responses of structures which are definable by material was given in section 3.8. The components discussed here are field effect transistors, transducers, temperature sensors, magnetic components, superconductors, mechanical sensors, and miscellaneous electronic components

Material and component progress within ARCHER for advanced high temperature reactor

International Nuclear Information System (INIS)

Buckthorpe, D.E.; Davies, M.; Pra, F.; Bonnamy, P.; Fokkens, J.; Heijna, M.; Bout, N. de; Vreeling, A.; Bourlier, F.; Lhachemi, D.; Woayehune, A.; Dubiez-le-Goff, S.; Hahner, P.; Futterer, M.; Berka, J.; Kalivodora, J.; Pouchon, M.A.; Schmitt, R.; Homerin, P.; Marsden, B.; Mummery, P.; Mutch, G.; Ponca, D.; Buhl, P.; Hoffmann, M.; Rondet, F.; Pecherty, A.; Baurand, F.; Alenda, F.; Esch, M.; Kohlz, N.; Reed, J.; Fachinger, J.; Klower, Dr.

2014-01-01

The ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R and D) integrated project started in 2011 as part of the European Commission 7. Framework Programme (FP7) for a period of four years to perform High Temperature Reactor technology R and D in support of reactor demonstration. The project consortium encompasses conventional and Nuclear Industry, Utilities, Technical Support Organizations, Research and Development Organizations and Academia. The activities involved contribute to the Generation IV (GIF) International Forum and collaborate with related projects in the US, China, Japan, and the Republic of Korea in cooperation with IAEA and ISTC. This paper addresses the progress of the work on materials and component technologies within ARCHER over the first two years of the project. (authors)

Laser Welding of Sub-assemblies before Forming

DEFF Research Database (Denmark)

Rasmussen, Mads; Olsen, Flemmming Ove; Pecas, Paulo

1996-01-01

This paper describes some experimental investigations of the formability of CO2-laser-welded 0.75 mm and 1.25 mm low carbon steel. There will be a description of how the laser welded blanks behave in different forming tests, and the influene of misalignment and undercut on the formability....... The quality is evalutated by measuring the imit strain and the limit effective strain for the laser welded sheets and the base material. These strains will be presented in a forming limit diagram (FLD). Finally the formability of the laser sheets is compared to that of the base materials....

On the applicability of probabilistic analyses to assess the structural reliability of materials and components for solid-oxide fuel cells

Energy Technology Data Exchange (ETDEWEB)

Lara-Curzio, Edgar [ORNL; Radovic, Miladin [Texas A& M University; Luttrell, Claire R [ORNL

2016-01-01

The applicability of probabilistic analyses to assess the structural reliability of materials and components for solid-oxide fuel cells (SOFC) is investigated by measuring the failure rate of Ni-YSZ when subjected to a temperature gradient and comparing it with that predicted using the Ceramics Analysis and Reliability Evaluation of Structures (CARES) code. The use of a temperature gradient to induce stresses was chosen because temperature gradients resulting from gas flow patterns generate stresses during SOFC operation that are the likely to control the structural reliability of cell components The magnitude of the predicted failure rate was found to be comparable to that determined experimentally, which suggests that such probabilistic analyses are appropriate for predicting the structural reliability of materials and components for SOFCs. Considerations for performing more comprehensive studies are discussed.

SALSA-A new instrument for strain imaging in engineering materials and components

International Nuclear Information System (INIS)

Pirling, Thilo; Bruno, Giovanni; Withers, Philip J.

2006-01-01

Residual stresses are very hard to predict and if undetected can lead to premature failure or unexpected behaviour of engineering materials or components. This paper describes the operation of a new residual strain-mapping instrument, Strain Analyser for Large and Small scale engineering Applications (SALSA), recently commissioned at the public user facility, the Institut Laue-Langevin in Grenoble, France. A unique feature of this neutron diffraction instrument is the sample manipulator, which is the first of its kind, allowing precise scanning of large and heavy (<500 kg) samples along any trajectory involving translations, tilts and rotations. Other notable features of the instrument are also described

Fabrication and Characterizations of Materials and Components for Intermediate Temperature Fuel Cells and Water Electrolysers

DEFF Research Database (Denmark)

Jensen, Annemette Hindhede; Prag, Carsten Brorson; Li, Qingfeng

The worldwide development of fuel cells and electrolysers has so far almost exclusively addressed either the low temperature window (20-200 °C) or the high temperature window (600-1000 °C). This work concerns the development of key materials and components of a new generation of fuel cells...... and electrolysers for operation in the intermediate temperature range from 200 to 400 °C. The intermediate temperature interval is of importance for the use of renewable fuels. Furthermore electrode kinetics is significantly enhanced compared to when operating at low temperature. Thus non-noble metal catalysts...... might be used. One of the key materials in the fuel cell and electrolyser systems is the electrolyte. Proton conducting materials such as cesium hydrogen phosphates, zirconium hydrogen phosphates and tin pyrophosphates have been investigated by others and have shown interesting potential....

Fast plasma sintering delivers functional graded materials components with macroporous structures and osseointegration properties.

Science.gov (United States)

Godoy, R F; Coathup, M J; Blunn, G W; Alves, A L; Robotti, P; Goodship, A E

2016-04-13

We explored the osseointegration potential of two macroporous titanium surfaces obtained using fast plasma sintering (FPS): Ti macroporous structures with 400-600 µmØ pores (TiMac400) and 850-1000 µmØ pores (TiMac850). They were compared against two surfaces currently in clinical use: Ti-Growth® and air plasma spray (Ti-Y367). Each surface was tested, once placed over a Ti-alloy and once onto a CoCr bulk substrate. Implants were placed in medial femoral condyles in 24 sheep. Samples were explanted at four and eight weeks after surgery. Push-out loads were measured using a material-testing system. Bone contact and ingrowth were assessed by histomorphometry and SEM and EDX analyses. Histology showed early osseointegration for all the surfaces tested. At 8 weeks, TiMac400, TiMac850 and Ti-Growth® showed deep bone ingrowth and extended colonisation with newly formed bone. The mechanical push-out force was equal in all tested surfaces. Plasma spray surfaces showed greater bone-implant contact and higher level of pores colonisation with new bone than FPS produced surfaces. However, the void pore area in FPS specimens was significantly higher, yet the FPS porous surfaces allowed a deeper osseointegration of bone to implant. FPS manufactured specimens showed similar osseointegration potential to the plasma spray surfaces for orthopaedic implants. FPS is a useful technology for manufacturing macroporous titanium surfaces. Furthermore, its capability to combine two implantable materials, using bulk CoCr with macroporous titanium surfaces, could be of interest as it enables designers to conceive and manufacture innovative components. FPS delivers functional graded materials components with macroporous structures optimised for osseointegration.

Consequences of the improvement of fast reactor material behavior under irradiation on fuel element performance

International Nuclear Information System (INIS)

Leclere, J.; Dupouy, J.M.; Marcon, J.P.

1979-01-01

The most important problems in fast reactor fuel element come from the excessive swelling of the structural materials used. The limitations of irradiation time for a given reactor result from the cladding or hexagonal wrapper deformations. Irradiation creep plays a major role, either in inducing additional deformations, or in providing possible ways of accommodation of bending stresses. Progress has been made in designing swelling resistant and/or low irradiation creep modulus materials. For instance in FRANCE, annealed 316 SS has been eliminated from pin and subassembly, and replaced by cold worked 316; we are now considering introduction of stabilizing elements in 316 SS as a further improvement and studying different alloys (nickel alloys, or ferritic steels). It has to be checked that the improvement of irradiation characteristic is not counterbalanced by losses on other properties (embrittlement for instance). Considering that pushing off or eliminating a limit may lead to the onset of a new one, it is porposed to make a review of the consequences of substantial improvement of structural material behavior

Flowing and freezing of molten core materials during unprotected loss of flow accidents in sodium cooled fast reactors

International Nuclear Information System (INIS)

Maschek, W.; Royl, P.

1988-09-01

Flowing and freezing of mobile core materials change the fissile material distribution and core-inventory under hypothetical accident conditions and determine the path to permanent shutdown of the neutronic events and the energetic potentials. The report classifies the bondary conditions for such flowing and freezing processes by going through the different situations under which these processes can occur in the scenario of the unprotected loss of flow (ULOF) accident. The classification is based on ULOF-accident simulations for a homogeneous reactor core concept of a 300 MWe LMFBR (e. g. SNR-300), but many boundary conditions are also characteristic for other core designs. A review of the relevant experiments is then made to correlate the available experimental information with these classified boundary conditions and to look at the resulting flowing and freezing processes. Boundary conditions that have been experimentally shown to be important are assigned high priorities. The data are specifically valued in relation to these boundary conditions of high priorities. The review includes the major experimental programs with published results. The discussion shows that the results from most clean condition tests for melt relocations are valuable for a better understanding of basic phenomena and analytical model development, but are not directly applicable to real accident conditions. The database for relevant boundary conditions from the ULOF scenario is limited and largely included in integral sequence tests from which quantitative information for modelling is difficult to obtain. Needs for additional investigations are identified. The suggestions are mainly restricted to investigations of the early phase of fuel removal. They are given with reference to candidate facilities and include relocations in the subassemblies and in the inter-subassembly gaps. Particular emphasis is put on the leading edge properties and possible driving forces to which more attention

Determination of basic components and oxidation power of Bi-Sr-Ca-Cu-O high temperature superconducting materials

International Nuclear Information System (INIS)

Vesene, T.B.

1993-01-01

A combination of methods is suggested for photometric determination of basic components of samples of bismuth-, strontium-, calcium-, and copper-based materials. A microchemical analysis is performed from one specimen without preliminary separation of components. Bismuth is singled out in the form of its complex with xylenol orange, calcium is defined in the form if its complex with calcion, strontium - in the form of its complex with chlorophosphonazo 3, and copper - in the form of its complex with PAR. Nonstoichiometric oxygen is detected via photometric evaluation of triiodine ions

Planning for closure and deactivation of the EBR-II complex

International Nuclear Information System (INIS)

Michelbacher, J.A.; Henslee, S.P.; Poland, H.F.; Wells, P.B.

1997-01-01

In January 1994, DOE terminated the Integral Fast Reactor (IFR) Program. Argonne National Laboratory-West (ANL-W) prepared a detailed plan to put Experimental Breeder Reactor-II (EBR-II) in a safe condition, including removal of irradiated fueled subassemblies from the plant, transfer of subassemblies, and removal and stabilization of primary and secondary sodium liquid heat transfer metal. The goal of deactivation is to stabilize the EBR-II complex until decontamination and decommissioning (D ampersand D) is implemented, thereby minimizing maintenance and surveillance. Deactivation of a sodium cooled reactor presents unique concerns. Residual sodium in the primary and secondary systems must be either reacted or inerted to preclude concerns with explosive sodium-air reactions. Also, residual sodium on components will effectively solder these items in place, making removal unfeasible. Several special cases reside in the primary system, including primary cold traps, a cesium trap, a cover gas condenser, and systems containing sodium-potassium alloy. The sodium or sodium-potassium alloy in these components must be reacted in place or the components removed. The Sodium Components Maintenance Shop at ANL-W provides the capability for washing primary components, removing residual quantities of sodium while providing some decontamination capacity. Considerations need to be given to component removal necessary for providing access to primary tank internals for D ampersand D activities, removal of hazardous materials, and removal of stored energy sources. ANL-W's plan for the deactivation of EBR-II addresses these issues, providing for an industrially and radiologically safe complex, requiring minimal surveillance during the interim period between deactivation and D ampersand D. Throughout the deactivation and closure of the EBR-II complex, federal environmental concerns will be addressed, including obtaining the proper permits for facility condition and waste processing

Report on the joint meeting of the Division of Development and Technology Plasma/Wall Interaction and High Heat Flux Materials and Components Task Groups

International Nuclear Information System (INIS)

Wilson, K.L.

1985-10-01

This report of the Joint Meeting of the Division of Development and Technology Plasma/Wall Interaction and High Heat Flux Materials and Components Task Groups contains contributing papers in the following areas: Plasma/Materials Interaction Program and Technical Assessment, High Heat Flux Materials and Components Program and Technical Assessment, Pumped Limiters, Ignition Devices, Program Planning Activities, Compact High Power Density Reactor Requirements, Steady State Tokamaks, and Tritium Plasma Experiments. All these areas involve the consideration of High Heat Flux on Materials and the Interaction of the Plasma with the First Wall. Many of the Test Facilities are described as well

CORROSION ISSUES ASSOCIATED WITH AUSTENITIC STAINLESS STEEL COMPONENTS USED IN NUCLEAR MATERIALS EXTRACTION AND SEPARATION PROCESSES

Energy Technology Data Exchange (ETDEWEB)

Mickalonis, J.; Louthan, M.; Sindelar, R.

2012-12-17

This paper illustrated the magnitude of the systems, structures and components used at the Savannah River Site for nuclear materials extraction and separation processes. Corrosion issues, including stress corrosion cracking, pitting, crevice corrosion and other corrosion induced degradation processes are discussed and corrosion mitigation strategies such as a chloride exclusion program and corrosion release testing are also discussed.

Laser welding of tailored blanks

International Nuclear Information System (INIS)

Pecas, P.; Gouveia, H.; Quintino, L.; Olsen, F.O.

1998-01-01

Laser welding has an increasing role in the automotive industry, namely on the sub-assemblies manufacturing. Several sheet-shape parts are laser welded, on a dissimilar combination of thicknesses and materials, and are afterwards formed (stamped) being transformed in a vehicle body component. In this paper low carbon CO 2 laser welding, on the thicknesses of 1,25 and 0.75 mm, formability investigation is described. There will be a description of how the laser welded blanks behave in different forming tests, and the influence of misalignment and undercut on the formability. The quality is evaluated by measuring the limit strain and limit effective strain for the laser welded sheets and the base material, which will be presented in a forming limit diagram. (Author) 14 refs

The use of boiling noise detection as a protection against faults in sub-assemblies in LMFBRs. Status report of work in the United Kingdom

International Nuclear Information System (INIS)

Burton, E.J.; MacLeod, I.D.

1982-01-01

The development of acoustic techniques for the surveillance of LMFBRs has the objective of providing a monitoring system on-line to give an early warning of incipient failures whilst the reactor is at power at present in the UK. Most attention is being given to safety protection to meet the design proposals for the Commercial Demonstration Fast Reactor (CDFR). One concern in the safety analysis is the hypothetical possibility that a local fault in a subassembly, if undetected could spread to its neighbours, eventually involving the whole core. An early warning of such a potentially propagating event would be given by detecting the boiling of the sodium. The specification of the acoustic technique, and therefore of the development programme, is set by the requirements of the safety analysis and the important features are outlined in the first section of the paper. This is followed by a description of the signal strength from boiling, based on out-of-pile experiments. This signal hat to be discriminated against the background noise arising from thc coolant pumps and the subassembly gag and flow noise. The detection of the acoustic signal may now be made by transducers rather than waveguides provided that the transducers are shown to be reliable enough and the recent work is summarised in the next section. The estimate of the signal/noise ratio depends upon the. transmission of the acoustic waves through the core to the sensor position. There is little experience on transmission in the reactor environment, possibilities for experiments are limited and laboratory tests are being used to improve basic knowledge. Modern computers offer the possibility of improving the sensitivity of detection by advanced data processing and the techniques which are being pursued are briefly described. Although acoustic technology has made great improvements in the last decade, especially in the application of acoustic emission techniques in thermal reactors, there is no experience of the

The use of boiling noise detection as a protection against faults in sub-assemblies in LMFBRs. Status report of work in the United Kingdom

Energy Technology Data Exchange (ETDEWEB)

Burton, E J; MacLeod, I D [United Kingdom Atomic Energy Authority, Risley Nuclear Power Development Laboratories, Risley, Warrington (United Kingdom)

1982-01-01

The development of acoustic techniques for the surveillance of LMFBRs has the objective of providing a monitoring system on-line to give an early warning of incipient failures whilst the reactor is at power at present in the UK. Most attention is being given to safety protection to meet the design proposals for the Commercial Demonstration Fast Reactor (CDFR). One concern in the safety analysis is the hypothetical possibility that a local fault in a subassembly, if undetected could spread to its neighbours, eventually involving the whole core. An early warning of such a potentially propagating event would be given by detecting the boiling of the sodium. The specification of the acoustic technique, and therefore of the development programme, is set by the requirements of the safety analysis and the important features are outlined in the first section of the paper. This is followed by a description of the signal strength from boiling, based on out-of-pile experiments. This signal had to be discriminated against the background noise arising from thc coolant pumps and the subassembly gag and flow noise. The detection of the acoustic signal may now be made by transducers rather than waveguides provided that the transducers are shown to be reliable enough and the recent work is summarised in the next section. The estimate of the signal/noise ratio depends upon the transmission of the acoustic waves through the core to the sensor position. There is little experience on transmission in the reactor environment, possibilities for experiments are limited and laboratory tests are being used to improve basic knowledge. Modern computers offer the possibility of improving the sensitivity of detection by advanced data processing and the techniques which are being pursued are briefly described. Although acoustic technology has made great improvements in the last decade, especially in the application of acoustic emission techniques in thermal reactors, there is no experience of the

Analysis on the Industrial Design of Food Package and the Component of Hazardous Substance in the Packaging Material

OpenAIRE

Wei-Wen Huang

2015-01-01

Transferring the hazardous chemicals contained in food packaging materials into food would threaten the health of consumers, therefore, the related laws and regulations and the detection method of hazardous substance have been established at home and abroad to ensure the safety to use the food packaging material. According to the analysis on the hazardous component in the food packaging, a set of detection methods for hazardous substance in the food packaging was established in the paper and ...

Interlocking multi-material components made of structured steel sheets and high-pressure die cast aluminium

Science.gov (United States)

Senge, S.; Brachmann, J.; Hirt, G.; Bührig-Polaczek, A.

2017-10-01

Lightweight design is a major driving force of innovation, especially in the automotive industry. Using hybrid components made of two or more different materials is one approach to reduce the vehicles weight and decrease fuel consumption. As a possible way to increase the stiffness of multi-material components, this paper presents a process chain to produce such components made of steel sheets and high-pressure die cast aluminium. Prior to the casting sequence the steel sheets are structured in a modified rolling process which enables continuous interlocking with the aluminium. Two structures manufactured by this rolling process are tested. The first one is a channel like structure and the second one is a channel like structure with undercuts. These undercuts enable the formation of small anchors when the molten aluminium fills them. The correlation between thickness reduction during rolling and the shape of the resulting structure was evaluated for both structures. It can be stated that channels with a depth of up to 0.5 mm and a width of 1 mm could be created. Undercuts with different size depending on the thickness reduction could be realised. Subsequent aluminium high-pressure die casting experiments were performed to determine if the surface structure can be filled gap-free with molten aluminium during the casting sequence and if a gap-free connection can be achieved after contraction of the aluminium. The casting experiments showed that both structures could be filled during the high-pressure die casting. The channel like structure results in a gap between steel and aluminium after contraction of the cast metal whereas the structure with undercuts leads to a good interlocking resulting in a gap-free connection.

e-Commerce and supply chains: Modelling of dynamics through ...

Indian Academy of Sciences (India)

R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

The dynamics associated with two production planning and control policies are modelled, viz. .... Hence, there is a strong need to design a dynamic knowledge inference system .... sell a variety of components to the subassembly manufacturer.

Fusion-component lifetime analysis

International Nuclear Information System (INIS)

Mattas, R.F.

1982-09-01

A one-dimensional computer code has been developed to examine the lifetime of first-wall and impurity-control components. The code incorporates the operating and design parameters, the material characteristics, and the appropriate failure criteria for the individual components. The major emphasis of the modeling effort has been to calculate the temperature-stress-strain-radiation effects history of a component so that the synergystic effects between sputtering erosion, swelling, creep, fatigue, and crack growth can be examined. The general forms of the property equations are the same for all materials in order to provide the greatest flexibility for materials selection in the code. The individual coefficients within the equations are different for each material. The code is capable of determining the behavior of a plate, composed of either a single or dual material structure, that is either totally constrained or constrained from bending but not from expansion. The code has been utilized to analyze the first walls for FED/INTOR and DEMO and to analyze the limiter for FED/INTOR

A guide to the effects of nuclear irradiation on materials, electronic and electrical components, oils and greases

International Nuclear Information System (INIS)

Lewis, R.C.E.

1983-03-01

A review is given of the effects of radiation on a selection of materials. Potentially damaging radiation thresholds are listed. The information has been taken from 12 documents for which references are given. The results of a study comparing component fault rates for nuclear and non-nuclear chemical plant are presented. (U.K.)

Radiation-induced structural transitions in composite materials with strong interaction of polymer components

International Nuclear Information System (INIS)

Zaikin, Yu.A.; Koztaeva, U.P.

2002-01-01

In earlier papers the internal friction (IF) method was applied to studies of structural relaxation in different types of polymer-based composite materials (glass-cloth, paper-based and foiled laminates impregnated by epoxy and phenolic resins) irradiated by 2 MeV electrons in the dose range of 0.1-50.0 MGy. Selectivity and high sensibility of the internal friction method allowed to distinguish glassy transitions in different structural components of the composites. The relaxation processes observed were identified and attributed to structural alterations in the polymer filler, the binder and the boundary layers. It was shown that changes in the parameters of relaxation maximums during irradiation can be considered as quantitative characteristics for the degree of radiation-induced degradation or cross-linking of polymer molecules. This paper deals with specific features of IF spectra in paper-based laminates where both the filler fibers and the binder are strongly interacting polymers. Anisotropy of viscous and elastic properties is very weak for this kind of materials, so that IF measurements give nearly the same result independently on the filler fiber orientation in the sample. The main reasons for it are the rigid chain structure of fillers (polyethylene-terephthalate and cellulose) and the good adhesion strengthened by diffusion of the epoxy or phenolic binder to defect regions of the filler.The IF temperature dependence observed in paper-based laminates is represented by superposition of two very broad relaxation maximums associated with transitions from glassy to high-elastic state in structural components, each based on one of the polymers. The inflection points characteristic for IF temperature dependence in paper-based laminates give a reason to treat them as a superposition of α-peaks associated with transitions from glassy to high-elastic state in structural components of a composite based on the binder and the filler, respectively. Another

Improvement of the material and transport component of the system of construction waste management

Science.gov (United States)

Kostyshak, Mikhail; Lunyakov, Mikhail

2017-10-01

Relevance of the topic of selected research is conditioned with the growth of construction operations and growth rates of construction and demolition wastes. This article considers modern approaches to the management of turnover of construction waste, sequence of reconstruction or demolition processes of the building, information flow of the complete cycle of turnover of construction and demolition waste, methods for improvement of the material and transport component of the construction waste management system. Performed analysis showed that mechanism of management of construction waste allows to increase efficiency and environmental safety of this branch and regions.

Implication of irradiation effects on materials data for the design of near core components

International Nuclear Information System (INIS)

Dietz, W.; Breitling, H.

1995-01-01

For LWR's strict regulations exist for the consideration of irradiation in the design and surveillance of the reactor pressure vessel in the various codes (ASME, RCC-M, KTA) but less for near core components. For FBR's no firm rules exist either for the vessel nor the reactor internals. In this paper the German design practices for the loop type SNR-300 will be presented, and also some information from the surveillance programme of the KNK-reactor. Austenitic stainless steels have been mainly selected for the near core components. For some special applications Ni-alloys and a stabilized 2 1/4 Cr 1 Mo-alloy were specified. Considerations of the irradiation effects on material properties will be made for the various temperature and fluence levels around the core. The surveillance programmes will be described. Both, the consideration of irradiation effects in the elastic and inelastic analysis and the surveillance programmes had been a part of the licensing process for SNR-300. (author). 8 figs, 4 tabs

Interactions between cask components and content of packaging for the transport of radioactive material during drop tests

International Nuclear Information System (INIS)

Quercetti, T.; Ballheimer, V.; Zeisler, P.; Mueller, K.

2003-01-01

This paper describes the analytical, numerical and experimental investigations on the phenomenon of interactions between cask components and content of packages for the transport of radioactive material during drop tests required according to the IAEA Regulations for the Safe Transport of Radioactive Material. Radial and axial gaps between cask components and content are usually necessary for thermal reasons but larger gaps can exist because of the geometrical dimensions of the specified content. Consequently interactions between content and cask components (lid system, cask body, etc.) are possible and can not be excluded during drop tests. Interactions in this context are relative movements between cask and content which are mainly due to elastic spring effects after releasing the cask for the free drop. These relative movements can cause interior collisions between content and cask during the main impact of the package onto the unyielding target. Drop tests with various types of Type A and Type B packages fully instrumented with strain gauges and accelerometers showed that these interactions respectively interior collisions can be considerable relating to high forces acting on cask lids, lid bolts and the content. Of course the real quantitative consequences of the interactions depend upon different conditions, among others the drop orientation, the design characteristics of the impact limiters, the dimensions of the gaps, the material characteristics of the contents, etc. . In order to investigate more precisely the phenomenon of interactions BAM carried out finite element calculations for the named casks using the ABAQUS/ Standard and ABAQUS/ Explicit computer code comparing them with results obtained from experiments. Additionally, tests with a simplified model instrumented with accelerometers were carried out accompanied by finite element calculations and analytical calculations using MATHEMATICA. The investigations on the mentioned phenomena of interaction

Correlation of creep and swelling with fuel pin performance

International Nuclear Information System (INIS)

Jackson, R.J.; Washburn, D.F.; Garner, F.A.; Gilbert, E.R.

1975-09-01

The HEDL PNL-11 experiment described was one in a series of fueled subassemblies irradiated in EBR-II to demonstrate the adequacy of the FFTF fuel pin design. The cladding material, dimensions, and fuel density are prototypic of FFTF. Because neutron flux in EBR-II is lower than in FFTF, the uranium enrichment is higher in these experimental fuel pins, irradiated in EBR-II, than the FFTF enrichment for comparable linear heat rates. Some pertinent oprating conditions for the center fuel pin in this experiment are listed. This 37-pin subassembly represents, at 110,000 MWd/MTM, the highest burnup yet attained by a prototypic FFTF subassembly. Similarly, this is the highest fluence presently attained by prototypic fuel pins. A cladding breach occurred in one fuel pin which is presently being examined. Results are presented and discussed

Recent Progress on the Key Materials and Components for Proton Exchange Membrane Fuel Cells in Vehicle Applications

Directory of Open Access Journals (Sweden)

Cheng Wang

2016-07-01

Full Text Available Fuel cells are the most clean and efficient power source for vehicles. In particular, proton exchange membrane fuel cells (PEMFCs are the most promising candidate for automobile applications due to their rapid start-up and low-temperature operation. Through extensive global research efforts in the latest decade, the performance of PEMFCs, including energy efficiency, volumetric and mass power density, and low temperature startup ability, have achieved significant breakthroughs. In 2014, fuel cell powered vehicles were introduced into the market by several prominent vehicle companies. However, the low durability and high cost of PEMFC systems are still the main obstacles for large-scale industrialization of this technology. The key materials and components used in PEMFCs greatly affect their durability and cost. In this review, the technical progress of key materials and components for PEMFCs has been summarized and critically discussed, including topics such as the membrane, catalyst layer, gas diffusion layer, and bipolar plate. The development of high-durability processing technologies is also introduced. Finally, this review is concluded with personal perspectives on the future research directions of this area.

Experience on sodium removal from FBTR components in its operating phase

International Nuclear Information System (INIS)

Jambunathan, D.; Rao, M.S.; Krishnamachari, V.S.; Kasiviswanathan, K.V.; Rajan, M.

1997-01-01

FBTR is a 40 MWt/13 MWe loop type, sodium cooled mixed carbide fuelled reactor. There are two primary loops, two secondary loops and a common steam-water circuit. Criticality was achieved in 1985 and during the course of the 10 years of operation phase experience has been gained on the decontamination of certain core components, primary sodium pumps, CRDM parts, handling components and cold traps. This paper deals with the decontamination aspects of these components in detail. For core subassemblies a remote sodium cleaning system was provided in the hot cell. The unique feature of this system is that it pumps alcohol into wash chamber using compressed nitrogen. During cleaning, the sub assembly is loaded into the system using master slave manipulator and compressed nitrogen gas is used to pump alcohol into the system for cleaning the outer and inner surfaces of the sub assembly in cycles. Each cycle takes about 15 minutes and around 5 to 6 cycles of cleaning were employed to remove 100 g of sodium. The level of alcohol in the tank is measured by ultrasonic level probes. The used alcohol is pumped to medium active waste storage tank. Primary and secondary sodium pumps are the large components which were cleaned off sodium using steam and water in decontamination pit. Lower part of CRDM with a failed bellow was another component cleaned in decontamination pit. An electro decontamination technique was successfully developed to remove coloration on the lower part of CRDM for reuse. A stainless steel carrier with antimony capsule was the first radio active component to undergo sodium cleaning operation in decontamination pit meant for large primary sodium circuit components after making necessary modifications. Decontamination of other components such as fingers of grippers and scrapper rings of charging and discharging flasks was carried out with alcohol under inert atmosphere. The secondary loop cold trap was successfully cleaned by hydride decomposition and vacuum

Development of low friction materials for LMFBR components

International Nuclear Information System (INIS)

Johnson, R.N.; Aungst, R.C.; Hoffman, N.J.; Cowgill, M.G.; Whitlow, G.A.; Wilson, W.L.

1976-01-01

The number of materials capable of providing low friction, low wear, and good corrosion resistance in low-oxygen (less than 1 ppM) sodium at temperatures up to 650 0 C are extremely limited. The paper describes the development, evaluation, and qualification of low-friction materials for this environment with emphasis on chromium carbide base coatings and nickel aluminide diffusion coatings. Design criteria and typical applications in liquid-metal-cooled reactors are described and recommendations offered for conditions under which these materials should and, perhaps more importantly, should not be used. Design parameters required to achieve optimum performance of these materials are discussed

A novel method for in-situ estimation of time constant for core temperature monitoring thermocouples of operating reactors

International Nuclear Information System (INIS)

Sylvia, J.I.; Chandar, S. Clement Ravi; Velusamy, K.

2014-01-01

Highlights: • Core temperature sensor was mathematically modeled. • Ramp signal generated during reactor operating condition is used. • Procedure and methodology has been demonstrated by applying it to FBTR. • Same technique will be implemented for all fast reactors. - Abstract: Core temperature monitoring system is an important component of reactor protection system in the current generation fast reactors. In this system, multiple thermocouples are housed inside a thermowell of fuel subassemblies. Response time of the thermocouple assembly forms an important input for safety analysis of fast reactor and hence frequent calibration/time constant estimation is essential. In fast reactors the central fuel subassembly is provided with bare fast response thermocouples to detect under cooling events in reactor and take proper safety action. On the other hand, thermocouples in thermowell are mainly used for blockage detection in individual fuel subassemblies. The time constant of thermocouples in thermowell can drift due to creep, vibration and thermal fatigue of the thermowell assembly. A novel method for in-situ estimation of time constant is proposed. This method uses the Safety Control Rod Accelerated Mechanism (SCRAM) or lowering of control Rod (LOR) signals of the reactor along with response of the central subassembly thermocouples as reference data. Validation of the procedure has been demonstrated by applying it to FBTR

Environmental Assessment for DOE permission for off-loading activities to support the movement of Millstone Unit 2 steam generator sub-assemblies across the Savannah River Site

International Nuclear Information System (INIS)

1992-10-01

The Department of Energy (DOE) has prepared an Environmental Assessment (EA), for the proposed granting of DOE permission of offloading activities to support the movement Millstone Unit 2 steam generator sub-assemblies (SGSAs) across the Savannah River Site (SRS). Based on the analyses in the EA, DOE has determined that the proposed action is not a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, an environmental impact statement is not required, and the Department is issuing this Finding of No Significant Impact. On the basis of the floodplain/wetlands assessment in the EA, DOE has determined that there is no practicable alternative to the proposed activities and that the proposed action has been designed to minimize potential harm to or within the floodplain of the SRS boat ramp. No wetlands on SRS would be affected by the proposed action

Materials Innovation for Next-Generation T&D Grid Components. Workshop Summary Report

Energy Technology Data Exchange (ETDEWEB)

Taylor, Emmanuel [Energetics Incorporated, Columbia, MD (United States); Kramer, Caroline [Energetics Incorporated, Columbia, MD (United States); Marchionini, Brian [Energetics Incorporated, Columbia, MD (United States); Sabouni, Ridah [Energetics Incorporated, Columbia, MD (United States); Cheung, Kerry [U.S. Department of Energy (DOE), Washington, DC (United States). Office of Electricity Delivery and Energy Reliability (OE); Lee, Dominic F [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-10-01

The Materials Innovations for Next-Generation T&D Grid Components Workshop was co-sponsored by the U.S. Department of Energy (DOE) Office of Electricity Delivery and Energy Reliability and the Oak Ridge National Laboratory (ORNL) and held on August 26 27, 2015, at the ORNL campus in Oak Ridge, Tennessee. The workshop was planned and executed under the direction of workshop co-chair Dr. Kerry Cheung (DOE) and co-chair Dr. Dominic Lee (ORNL). The information contained herein is based on the results of the workshop, which was attended by nearly 50 experts from government, industry, and academia. The research needs and pathways described in this report reflect the expert opinions of workshop participants, but they are not intended to represent the views of the entire electric power community.

A case of industrial safety appraisal for extension of service life of GTK-10-4 gas turbines used at gas transmission stations

Science.gov (United States)

Rybnikov, A. I.; Kovalev, A. G.; Kryukov, I. I.; Leont'ev, S. A.; Moshnikov, A. V.

2017-04-01

It is shown that the extended life and enhanced operational reliability of parts and subassemblies of the most popular GTK-10-4 gas transmission plants are determined by the enhanced efficiency of the control over technical condition and operational safety of turbine plants in conformity with industrial safety requirements imposed on gas pipeline compressor stations. It has been established that the materials of parts and subassemblies of gas turbine plants with different, especially with maximal operating time, shall be exposed to NDT for the purpose of determining the actual mechanical characteristics of these materials with different operating time and calculating residual life. The analysis of damageability and operating conditions has helped to identify parts and subassemblies for repair or replacement with the highest frequency of unacceptable defects. These parts and subassemblies have been shown to include base members of the axial compressor (AC), a turbine housing, an axial compressor rotor, high- and low-pressure turbine (HPT and LPT) discs, a 12-part holder, the housing of the holder of HPT and LPT guiding blades, a sealed baffler, and working and guiding AC, LPT and HPT blades. The most typical operational defects have been enumerated and analyzed. It has been determined that the primary task of the industrial safety appraisal for extending the life of GTK-10-4 with limit-exceeding operating time is to thoroughly examine HPT and LPT discs with more than 130,000 hours of operating time and establish by DT methods characteristics of materials for evaluation, taking account of their degradation, and residual life of critical turbine elements. In addition, it has been shown that the service life of HP turbine discs can be extended by replacing the disc material (EP-428 12% chromium steel) with a material with a higher linear expansion factor that somewhat exceeds the expansion factor of EI-893 nickel alloy used to melt out working blades.

Inspection of disposal canisters components

International Nuclear Information System (INIS)

Pitkaenen, J.

2013-12-01

This report presents the inspection techniques of disposal canister components. Manufacturing methods and a description of the defects related to different manufacturing methods are described briefly. The defect types form a basis for the design of non-destructive testing because the defect types, which occur in the inspected components, affect to choice of inspection methods. The canister components are to nodular cast iron insert, steel lid, lid screw, metal gasket, copper tube with integrated or separate bottom, and copper lid. The inspection of copper material is challenging due to the anisotropic properties of the material and local changes in the grain size of the copper material. The cast iron insert has some acoustical material property variation (attenuation, velocity changes, scattering properties), which make the ultrasonic inspection demanding from calibration point of view. Mainly three different methods are used for inspection. Ultrasonic testing technique is used for inspection of volume, eddy current technique, for copper components only, and visual testing technique are used for inspection of the surface and near surface area

Advanced computational simulation for design and manufacturing of lightweight material components for automotive applications

Energy Technology Data Exchange (ETDEWEB)

Simunovic, S.; Aramayo, G.A.; Zacharia, T. [Oak Ridge National Lab., TN (United States); Toridis, T.G. [George Washington Univ., Washington, DC (United States); Bandak, F.; Ragland, C.L. [Dept. of Transportation, Washington, DC (United States)

1997-04-01

Computational vehicle models for the analysis of lightweight material performance in automobiles have been developed through collaboration between Oak Ridge National Laboratory, the National Highway Transportation Safety Administration, and George Washington University. The vehicle models have been verified against experimental data obtained from vehicle collisions. The crashed vehicles were analyzed, and the main impact energy dissipation mechanisms were identified and characterized. Important structural parts were extracted and digitized and directly compared with simulation results. High-performance computing played a key role in the model development because it allowed for rapid computational simulations and model modifications. The deformation of the computational model shows a very good agreement with the experiments. This report documents the modifications made to the computational model and relates them to the observations and findings on the test vehicle. Procedural guidelines are also provided that the authors believe need to be followed to create realistic models of passenger vehicles that could be used to evaluate the performance of lightweight materials in automotive structural components.

Chemical and biotechnological processing of collagen-containing raw materials into functional components of feed suitable for production of high-quality meat from farm animals

Science.gov (United States)

Baburina, M. I.; Ivankin, A. N.; Stanovova, I. A.

2017-09-01

The process of chemical biotechnological processing of collagen-containing raw materials into functional components of feeds for effective pig rearing was studied. Protein components of feeds were obtained as a result of hydrolysis in the presence of lactic acid of the animal collagen from secondary raw materials, which comprised subcutaneous collagen (cuticle), skin and veined mass with tendons from cattle. For comparison, a method is described for preparing protein components of feeds by cultivating Lactobacillus plantarum. Analysis of the kinetic data of the conversion of a high-molecular collagen protein to an aminolyte polypeptide mixture showed the advantage of microbiological synthesis in obtaining a protein for feeds. Feed formulations have been developed to include the components obtained, and which result in high quality pork suitable for the production of quality meat products.

Helium leak testing of large pressure vessels or subassemblies

International Nuclear Information System (INIS)

Hopkins, J.S.; Valania, J.J.

1977-01-01

Specifications for pressure-vessel components [such as the intermediate heat exchangers (IHX)] for service in the liquid metal fast breeder reactor facilities require helium leak testing of pressure boundaries to very exacting standards. The experience of Foster Wheeler Energy Corporation (FWEC) in successfully leak-testing the IHX shells and bundle assemblies now installed in the Fast Flux Test Facility at Richland, WA is described. Vessels of a somewhat smaller size for the closed loop heat exchanger system in the Fast Flux Test Facility have also been fabricated and helium leak tested for integrity of the pressure boundary by FWEC. Specifications on future components call for helium leak testing of the tube to tubesheet welds of the intermediate heat exchangers

Replaceable LMFBR core components

International Nuclear Information System (INIS)

Evans, E.A.; Cunningham, G.W.

1976-01-01

Much progress has been made in understanding material and component performance in the high temperature, fast neutron environment of the LMFBR. Current data have provided strong assurance that the initial core component lifetime objectives of FFTF and CRBR can be met. At the same time, this knowledge translates directly into the need for improved core designs that utilize improved materials and advanced fuels required to meet objectives of low doubling times and extended core component lifetimes. An industrial base for the manufacture of quality core components has been developed in the US, and all procurements for the first two core equivalents for FFTF will be completed this year. However, the problem of fabricating recycled plutonium while dramatically reducing fabrication costs, minimizing personnel exposure, and protecting public health and safety must be addressed

Experience with the instrumentation tests in large sodium test facilities

International Nuclear Information System (INIS)

Lauhoff, Th.; Ruppert, E.; Stehle, H.; Vinzens, K.

1976-01-01

A facility is described for fast breeder core components (AKB) to test specially instrumented fuel dummies and blanket elements, and also absorber elements under simulated normal and extreme reactor conditions. In addition to endurance testing of a special sodium and high temperature sub-assembly, instrumentation is provided to investigate thermohydraulic and vibrational behaviour of core elements. During tests of > 3000 h at temperatures above 820 K the main sub-assembly characteristics, e.g. pressure drop, leakage flow, vibration and noise spectra can be reproduced. The use of eddy current flow meters, strain gauges, magnetostrictive noise sensors, pressure transducers, thermocouples, and acoustic surveillance devices, are described. (U.K.)

MARTINS: A foam/film flow model for molten material relocation in HWRs with U-Al-fueled multi-tube assemblies

International Nuclear Information System (INIS)

Kalimullah.

1994-01-01

Some special purpose heavy-water reactors (EM) are made of assemblies consisting of a number of coaxial aluminum-clad U-Al alloy fuel tubes and an outer Al sleeve surrounding the fuel tubes. The heavy water coolant flows in the annular gaps between the circular tubes. Analysis of severe accidents in such reactors requires a model for predicting the behavior of the fuel tubes as they melt and disrupt. This paper describes a detailed, mechanistic model for fuel tube heatup, melting, freezing, and molten material relocation, called MARTINS (Melting and Relocation of Tubes in Nuclear subassembly). The paper presents the modeling of the phenomena in MARTINS, and an application of the model to analysis of a reactivity insertion accident. Some models are being developed to compute gradual downward relocation of molten material at decay-heat power levels via candling along intact tubes, neglecting coolant vapor hydrodynamic forces on molten material. These models are inadequate for high power accident sequences involving significant hydrodynamic forces. These forces are included in MARTINS

An anti-diffusive Lagrange-Remap scheme for multi-material compressible flows with an arbitrary number of components

Directory of Open Access Journals (Sweden)

Kokh Samuel

2012-04-01

Full Text Available We propose a method dedicated to the simulation of interface flows involving an arbitrary number m of compressible components. Our task is two-fold: we first introduce a m-component flow model that generalizes the two-material five-equation model of [2,3]. Then, we present a discretization strategy by means of a Lagrange-Remap [8,10] approach following the lines of [5,7,12]. The projection step involves an anti-dissipative mechanism derived from [11,12]. This feature allows to prevent the numerical diffusion of the material interfaces. We present two-dimensional simulation results of three-material flow. Nous proposons une méthode de simulation pour des écoulements comportant un nombre arbitraire m de composants compressibles séparés par des interfaces. Nous procdons en deux tapes : tout d’abord nous introduisons un modèle d’écoulementm composants qui généralise le modèle à cinq équations de [2,3]. Ensuite nous présentons une stratégie de discrétisation de type Lagrange-Projection [8,10] inspirée de [5,7,12]. La phase de projection met en œuvre une technique de transport anti-diffusive [11,12] qui permet de limiter la diffusion numérique des interfaces matérielles. Nous présentons des résultats de calcul bidimensionnel d’écoulement à trois composants.

Risk of false decision on conformity of a multicomponent material when test results of the components' content are correlated.

Science.gov (United States)

Kuselman, Ilya; Pennecchi, Francesca R; da Silva, Ricardo J N B; Hibbert, D Brynn

2017-11-01

The probability of a false decision on conformity of a multicomponent material due to measurement uncertainty is discussed when test results are correlated. Specification limits of the components' content of such a material generate a multivariate specification interval/domain. When true values of components' content and corresponding test results are modelled by multivariate distributions (e.g. by multivariate normal distributions), a total global risk of a false decision on the material conformity can be evaluated based on calculation of integrals of their joint probability density function. No transformation of the raw data is required for that. A total specific risk can be evaluated as the joint posterior cumulative function of true values of a specific batch or lot lying outside the multivariate specification domain, when the vector of test results, obtained for the lot, is inside this domain. It was shown, using a case study of four components under control in a drug, that the correlation influence on the risk value is not easily predictable. To assess this influence, the evaluated total risk values were compared with those calculated for independent test results and also with those assuming much stronger correlation than that observed. While the observed statistically significant correlation did not lead to a visible difference in the total risk values in comparison to the independent test results, the stronger correlation among the variables caused either the total risk decreasing or its increasing, depending on the actual values of the test results. Copyright © 2017 Elsevier B.V. All rights reserved.

Prediction of Corrosion of Advanced Materials and Fabricated Components

Energy Technology Data Exchange (ETDEWEB)

A. Anderko; G. Engelhardt; M.M. Lencka (OLI Systems Inc.); M.A. Jakab; G. Tormoen; N. Sridhar (Southwest Research Institute)

2007-09-29

The goal of this project is to provide materials engineers, chemical engineers and plant operators with a software tool that will enable them to predict localized corrosion of process equipment including fabricated components as well as base alloys. For design and revamp purposes, the software predicts the occurrence of localized corrosion as a function of environment chemistry and assists the user in selecting the optimum alloy for a given environment. For the operation of existing plants, the software enables the users to predict the remaining life of equipment and help in scheduling maintenance activities. This project combined fundamental understanding of mechanisms of corrosion with focused experimental results to predict the corrosion of advanced, base or fabricated, alloys in real-world environments encountered in the chemical industry. At the heart of this approach is the development of models that predict the fundamental parameters that control the occurrence of localized corrosion as a function of environmental conditions and alloy composition. The fundamental parameters that dictate the occurrence of localized corrosion are the corrosion and repassivation potentials. The program team, OLI Systems and Southwest Research Institute, has developed theoretical models for these parameters. These theoretical models have been applied to predict the occurrence of localized corrosion of base materials and heat-treated components in a variety of environments containing aggressive and non-aggressive species. As a result of this project, a comprehensive model has been established and extensively verified for predicting the occurrence of localized corrosion as a function of environment chemistry and temperature by calculating the corrosion and repassivation potentials.To support and calibrate the model, an experimental database has been developed to elucidate (1) the effects of various inhibiting species as well as aggressive species on localized corrosion of nickel

Tribological behaviour of skin equivalents and ex-vivo human skin against the material components of artificial turf in sliding

NARCIS (Netherlands)

Morales Hurtado, Marina; Peppelman, P.; Zeng, Xiangqiong; van Erp, P.E.J.; van der Heide, Emile

2016-01-01

This research aims to analyse the interaction of three artificial skin equivalents and human skin against the main material components of artificial turf. The tribological performance of Lorica, Silicone Skin L7350 and a recently developed Epidermal Skin Equivalent (ESE) were studied and compared to

Overview moderator material for nuclear reactor components

International Nuclear Information System (INIS)

Mairing Manutu Pongtuluran; Hendra Prihatnadi

2009-01-01

In order for a reactor design is considered acceptable absolute technical requirement is fulfilled because the most important part of a reactor design. Safety considerations emphasis on the handling of radioactive substances emitted during the operation of a reactor and radioactive waste handling. Moderator material is a layer that interacts directly with neutrons split the nuclear fuel that will lead to changes in physical properties, nuclear properties, mechanical properties and chemical properties. Reviews moderator of this time is of the types of moderator is often used to meet the requirements as nuclear material. (author)

Additive Manufacturing of Metallic and Ceramic Components by the Material Extrusion of Highly-Filled Polymers: A Review and Future Perspectives.

Science.gov (United States)

Gonzalez-Gutierrez, Joamin; Cano, Santiago; Schuschnigg, Stephan; Kukla, Christian; Sapkota, Janak; Holzer, Clemens

2018-05-18

Additive manufacturing (AM) is the fabrication of real three-dimensional objects from metals, ceramics, or plastics by adding material, usually as layers. There are several variants of AM; among them material extrusion (ME) is one of the most versatile and widely used. In MEAM, molten or viscous materials are pushed through an orifice and are selectively deposited as strands to form stacked layers and subsequently a three-dimensional object. The commonly used materials for MEAM are thermoplastic polymers and particulate composites; however, recently innovative formulations of highly-filled polymers (HP) with metals or ceramics have also been made available. MEAM with HP is an indirect process, which uses sacrificial polymeric binders to shape metallic and ceramic components. After removing the binder, the powder particles are fused together in a conventional sintering step. In this review the different types of MEAM techniques and relevant industrial approaches for the fabrication of metallic and ceramic components are described. The composition of certain HP binder systems and powders are presented; the methods of compounding and filament making HP are explained; the stages of shaping, debinding, and sintering are discussed; and finally a comparison of the parts produced via MEAM-HP with those produced via other manufacturing techniques is presented.

Additive Manufacturing of Metallic and Ceramic Components by the Material Extrusion of Highly-Filled Polymers: A Review and Future Perspectives

Science.gov (United States)

Cano, Santiago

2018-01-01

Additive manufacturing (AM) is the fabrication of real three-dimensional objects from metals, ceramics, or plastics by adding material, usually as layers. There are several variants of AM; among them material extrusion (ME) is one of the most versatile and widely used. In MEAM, molten or viscous materials are pushed through an orifice and are selectively deposited as strands to form stacked layers and subsequently a three-dimensional object. The commonly used materials for MEAM are thermoplastic polymers and particulate composites; however, recently innovative formulations of highly-filled polymers (HP) with metals or ceramics have also been made available. MEAM with HP is an indirect process, which uses sacrificial polymeric binders to shape metallic and ceramic components. After removing the binder, the powder particles are fused together in a conventional sintering step. In this review the different types of MEAM techniques and relevant industrial approaches for the fabrication of metallic and ceramic components are described. The composition of certain HP binder systems and powders are presented; the methods of compounding and filament making HP are explained; the stages of shaping, debinding, and sintering are discussed; and finally a comparison of the parts produced via MEAM-HP with those produced via other manufacturing techniques is presented. PMID:29783705

Increased component safety through improved methods for residual stress analysis. Subprojects. Consideration of the elastic-plastic material properties (phase 1). Final report

International Nuclear Information System (INIS)

Mirbach, David von

2014-01-01

Residual stresses in mechanical components can result in both detrimental but also beneficial effects on the strength and lifetime of the components. The most detailed knowledge of the residual stress state is of advantage or a pre-requisite for the assessment of the component performance. The mechanical methods for residual stress measurement are divided into the groups of non-destructive and destructive methods. Two commonly used mechanical methods for determination of residual stresses are the hole drilling method and the ring core method which can be regarded as semi-destructive methods. In the context of reactor safety research of the German Federal Ministry of Economic and Technology (BMWi) two fundamental and interacting weak points of the hole drilling method as well as of the ring core method, respectively, in order to determine residual stresses are going to be investigated. As a consequence reliability of the methods will be improved in this joint research project. On the one hand there are effects of geometrical boundary conditions of the components and on the other hand there is the influence of plasticity due to notch effects both affecting the released strain field after removing material and after all the calculated residual stresses. The first issue mentioned above is under the responsibility of the Institute of Materials Engineering (Kassel University) and the last one is investigated by Universitaet of Stuttgart-Otto-Graf-Institut - materials testing institute. As a consequence of a successful project the knowledge base will be considerably improved resulting in benefits for various engineering fields. Especially the quantitative consideration of real residual stress states for optimized component designs will be possible and after all the consequences of residual stresses on safety of components which are used in nuclear facilities can be evaluated. The state of art was reground in the first research chapter and the analysed strain gauges where

Development of a technology for amorphous material (Co-free) hardfacing on primary side component materials using laser beam to improve their wear/erosion.corrosion resistance

Energy Technology Data Exchange (ETDEWEB)

Suh, Jeong Hun; Kim, J. S.; Han, J. H.; Lee, D. H.; Hwang, S. S

2000-08-01

A technology of laser hardfacing of amorphous materials onto materials used in the primary-side components has been developed in order to improve their integrity and reduce the radiation fluence in the primary system. (1) Development of a powder feeding system for the laser cladding. (2) Modification of the laser system in order to perform cladding the part surfaces with complex 3D geometries through the tool paths determined with CAD/CAM. (3) Development of laser cladding technology with amorphous alloy. (4) Examination and analysis of the microstructure, chemical composition, and phases of the clads. (5) Evaluation of the mechanical properties of the clads. (6) Development of an ultrasonic vibrator for VSR.

Development of a technology for amorphous material (Co-free) hardfacing on primary side component materials using laser beam to improve their wear/erosion.corrosion resistance

International Nuclear Information System (INIS)

Suh, Jeong Hun; Kim, J. S.; Han, J. H.; Lee, D. H.; Hwang, S. S.

2000-08-01

A technology of laser hardfacing of amorphous materials onto materials used in the primary-side components has been developed in order to improve their integrity and reduce the radiation fluence in the primary system. 1) Development of a powder feeding system for the laser cladding. 2) Modification of the laser system in order to perform cladding the part surfaces with complex 3D geometries through the tool paths determined with CAD/CAM. 3) Development of laser cladding technology with amorphous alloy. 4) Examination and analysis of the microstructure, chemical composition, and phases of the clads. 5) Evaluation of the mechanical properties of the clads. 6) Development of an ultrasonic vibrator for VSR

Development of a technology for amorphous material (Co-free) hardfacing on primary side component materials using laser beam to improve their wear/erosion.corrosion resistance

International Nuclear Information System (INIS)

Suh, Jeong Hun; Kim, J. S.; Hwang, S. S.; Lim, Y. S.

1999-08-01

A technology of laser hardfacing of amorphous materials on materials used in the primary-side components has been developed in order to improve their integrity and reduce the radiation fluence in the primary system. 1) Development of a power feeding system for the primary system. 2) Modification of the laser system in order to perform cladding the part surfaces with complex 3D geometries through the tool paths determined with CAD/CAM. 3) Development of laser cladding technology with amorphous alloy. 4) Examination and analysis of the microstructure, chemical composition, and phase of the clad. 5) Evaluation of the mechanical properties of the clad. 6) Development of an ultrasonic vibrator for VSR. (author)

A research project to develop and evaluate a technical education component on materials technology for orientation to space-age technology

Science.gov (United States)

Jacobs, J. A.

1976-01-01

A project was initiated to develop, implement, and evaluate a prototype component for self-pacing, individualized instruction on basic materials science. Results of this project indicate that systematically developed, self-paced instruction provides an effective means for orienting nontraditional college students and secondary students, especially minorities, to both engineering technology and basic materials science. In addition, students using such a system gain greater chances for mastering subject matter than with conventional modes of instruction.

IFMIF [International Fusion Materials Irradiation Facility], an accelerator-based neutron source for fusion components irradiation testing: Materials testing capabilities

International Nuclear Information System (INIS)

Mann, F.M.

1988-08-01

The International Fusion Materials Irradiation Facility (IFMIF) is proposed as an advanced accelerator-based neutron source for high-flux irradiation testing of large-sized fusion reactor components. The facility would require only small extensions to existing accelerator and target technology originally developed for the Fusion Materials Irradiation Test (FMIT) facility. At the extended facility, neutrons would be produced by a 0.1-A beam of 35-MeV deuterons incident upon a liquid lithium target. The volume available for high-flux (>10/sup 15/ n/cm/sup 2/-s) testing in IFMITF would be over a liter, a factor of about three larger than in the FMIT facility. This is because the effective beam current of 35-MeV deuterons on target can be increased by a factor of ten to 1A or more. Such an increase can be accomplished by funneling beams of deuterium ions from the radio-frequency quadruple into a linear accelerator and by taking advantage of recent developments in accelerator technology. Multiple beams and large total current allow great variety in available testing. For example, multiple simultaneous experiments, and great flexibility in tailoring spatial distributions of flux and spectra can be achieved. 5 refs., 2 figs., 1 tab

To the application of TV and optical equipment for in-service inspection of reactor vessel and primary circuit component materials

International Nuclear Information System (INIS)

Afonin, Eh.M.; Bachelis, I.M.; Tokarev, E.A.; Yastrebov, V.E.

1985-01-01

Some problems of application of TV and optical equipment for inspection of reactor vessel and primary circuit component materials are considered taking the most widespread WWER-440 type reactor as an example. The most advanrageous objects of the inspection and typical zones of equipment arrangement are shown. Methods and peculiarities of the inspection with the use of TV and optical equipment are presented. Recommendations on rational application of the equipment for the inspection of WWER-440 reactor vessel components are given

Biomass-derived carbonaceous materials as components in wood briquettes

Energy Technology Data Exchange (ETDEWEB)

Stengl, S.; Koch, C.; Stadlbauer, E.A.; Scheer, J. [Univ. of Applied Sciences, THM Campus Giessen, Giessen (Germany); Weber, B. [Instituto de Ingenieria de la Universidad Nacional Autonoma de Mexico (UNAM), Coyoacan (Mexico); Strohal, U.; Fey, J. [Strohal Anlagenbau, Staufenberg (Germany)

2012-11-01

The present paper describes a briquette composed of a substantial amount of wooden biomass and up to 35% of carbonaceous materials derived from biogenic residues. The cellulosic component may be a mixture of any wooden residue. Suitable substrates for the carbonaceous fraction are vegetation wastes from land management or agriculture. Depending on physical and chemical nature of the substrate, Hydrothermal Carbonisation (HTC) or Low Temperature Conversion (LTC) may be used to produce the carbonaceous part of the briquette. HTC turns wet biomass at temperatures around 200 deg C in an autoclave into lignite whereas LTC treatment at 400 deg C and atmospheric pressure produces black coal. This is manifested by a molar ratio of 0.1 {<=} H/C (LTC) {<=} 0.7; 0.05{<=} O/C (LTC) {<=} 0.4 and 0.7 < H/C (HTC) <1.5 ; 0.2< O/C (HTC) < 0.5. Solid state {sup 13}C-NMR confirms these findings showing a strong absorption band for sp{sup 2}-hybridized carbon atoms at chemical shifts of 100 ppm und 165 ppm for LTC biochar. Depending on the substrate, HTC gives rise to an increase in the specific calorific value (MJ/kg) by a factor of {Psi} {approx} 1.2 - 1.4; LTC by 1.5 - 1.8. In addition ash melting points are significantly increased; in case of wheat straw by about 200 deg C. Compacted products may have a cylindrical or rectangular profile.

Turbine repair process, repaired coating, and repaired turbine component

Science.gov (United States)

Das, Rupak; Delvaux, John McConnell; Garcia-Crespo, Andres Jose

2015-11-03

A turbine repair process, a repaired coating, and a repaired turbine component are disclosed. The turbine repair process includes providing a turbine component having a higher-pressure region and a lower-pressure region, introducing particles into the higher-pressure region, and at least partially repairing an opening between the higher-pressure region and the lower-pressure region with at least one of the particles to form a repaired turbine component. The repaired coating includes a silicon material, a ceramic matrix composite material, and a repaired region having the silicon material deposited on and surrounded by the ceramic matrix composite material. The repaired turbine component a ceramic matrix composite layer and a repaired region having silicon material deposited on and surrounded by the ceramic matrix composite material.

Method for assembling dynamoelectric machine end shield parts

International Nuclear Information System (INIS)

Thomson, J.M.

1984-01-01

Methods, apparatus, and systems are provided for automatically assembling end shield assemblies of subassemblies for electric motors. In a preferred form, a system and methods are provided that utilize a non-palletized, non-synchronous concept to convey end shields through a number of assembly stations. At process stations situated along a conveyor, operations are performed on components. One method includes controlling traffic of sub-assemblies by toggle type escapements. A stop or latch of unique design stops end shield components in midstream, and ''lifts'' of unique design disengage parts from the conveyor and also support such parts during various operations. Photo-optic devices and proximity and reed switch mechanisms are utilized for control purposes. The work stations involved in one system include a unique assembly and pressing station involving oil well covers; a unique feed wick seating system; a unique lubricant adding operation; and unique ''building block'' mechanisms and methods

DNA as a component of ER materials

International Nuclear Information System (INIS)

Minagawa, K; Aoki, Y; Berber, M R; Mori, T; Tanaka, M

2009-01-01

Deoxyribonucleic acid (DNA), which is known as a typical biopolymer, has been utilized for a few types of ER materials. Suspensions were prepared with the particles of DNA, DNA/lipid complexes, and LDH (layered double hydroxide)/DNA composites. The purified DNA showed larger ER effect than the others, but this particle tended to absorb water, which caused less stability. Preliminary experiments of preparing composite with LDH indicated that this inorganic material would be useful for hydrophobic modification of DNA particles, although further optimization of composite preparation is needed. In addition, the LDH/DNA suspensions showed interesting behaviours under some conditions, which indicated possibility for controlling ER property in a wide range.

DNA as a component of ER materials

Energy Technology Data Exchange (ETDEWEB)

Minagawa, K; Aoki, Y; Berber, M R [Institute of Technology and Science, University of Tokushima, Tokushima 770-8506 (Japan); Mori, T [Department of Applied Chemistry, Faculty of Engineering, Kyushu University, Fukuoka 819-0395 (Japan); Tanaka, M [Faculty of Pharmaceutical Science, Tokushima Bunri University, Tokushima 770-8514 (Japan)], E-mail: minagawa@chem.tokushima-u.ac.jp

2009-02-01

Deoxyribonucleic acid (DNA), which is known as a typical biopolymer, has been utilized for a few types of ER materials. Suspensions were prepared with the particles of DNA, DNA/lipid complexes, and LDH (layered double hydroxide)/DNA composites. The purified DNA showed larger ER effect than the others, but this particle tended to absorb water, which caused less stability. Preliminary experiments of preparing composite with LDH indicated that this inorganic material would be useful for hydrophobic modification of DNA particles, although further optimization of composite preparation is needed. In addition, the LDH/DNA suspensions showed interesting behaviours under some conditions, which indicated possibility for controlling ER property in a wide range.

The effect of alkaline treatment and fiber orientation on impact resistant of bio-composites Sansevieria trifasciata fiber/polypropylene as automotive components material

Science.gov (United States)

Shieddieque, Apang Djafar; Mardiyati, Suratman, Rochim; Widyanto, Bambang

2018-04-01

The increasing amount of car usage is causing an escalated amount of fuel consumption and CO2 emission. It implicates demand for the automotive industry to increase the efficiency of their products, One of the most effective ways to solve the issue is to find green weight light material for the interior automotive component. The Aim of this research was to investigate the effect of alkaline treatment and fiber orientation on the impact resistant of material bio- composite sansevieiria trifasciata fiber/Polypropylene. In this research, bio-composites sansevieria trifasciata fiber/Polypropylene was prepared with random fiber orientation and unidirectional orientation by using a hot press method with pressure 140 Bar and temperature 240°C. Fiber was taken from Sansevieria trifasciata by using mechanical retting. In this study, Sansevieria fiber was given alkaline treatment (mercerization) with NaOH 3% (w/w) solution at temperature 100°C for an hour. The fraction of fiber volume that were used in this experiment are 0%, 5%, 10%, and 15%. The impact test was conducted based on ASTM D 6110 - 04, and the fracture analysis was investigated by scanning electron microscope (SEM). The best result of impact toughness and fracture analysis were achieved by bio composite untreated and unidirectional sansevieria trifasciata fiber/Polypropylene with fiber volume fraction of 15%, which was 48.092kJ/m2 for impact resistant. As compared to the impact toughness standard, which needed for interior automotive component, the impact toughness of sansevieria trifasciata fiber/Polypropylene has fulfilled the standard of the interior material automotive industry. Therefore, this material can be potentially used as materials on the car exterior component.

Report on the joint meeting of the Division of Development and Technology Plasma Wall Interaction and High Heat Flux Materials and Components task groups

International Nuclear Information System (INIS)

Nygren, R.E.

1992-04-01

The Plasma/Wall Interaction and High Heat Flux Materials and Components Task Groups typically hold a joint meeting each year to provide a forum for discussion of technical issues of current interest as well as an opportunity for program reviews by the Department of Energy (DOE). At the meeting in September 1990, reported here, research programs in support of the International Thermonuclear Experimental Reactor (ITER) were highlighted. The first part of the meeting was devoted to research and development (R ampersand D) for ITER on plasma facing components plus introductory presentations on some current projects and design studies. The balance of the meeting was devoted to program reviews, which included presentations by most of the participants in the Small Business Innovative Research (SBIR) Programs with activities related to plasma wall interactions. The Task Groups on Plasma/Wall Interaction and on High Heat Flux Materials and Components were chartered as continuing working groups by the Division of Development and Technology in DOE's Magnetic Fusion Program. This report is an addition to the series of ''blue cover'' reports on the Joint Meetings of the Plasma/Wall Interaction and High Heat Flux Materials and Components Task Groups. Among several preceding meetings were those in October 1989 and January 1988

Tower Shielding Reactor II design and operation report. Vol. 3. Assembling and testing of the control mechanism assembly

International Nuclear Information System (INIS)

Ward, D.R.; Holland, L.B.

1979-09-01

The mechanisms that are operated to control the reactivity of the Tower Shielding Reactor II(TSR-II) are mounted on a Control Mechanism Housing (CMH) that is centered inside the reactor core. The information required to procure, fabricate, inspect, and assemble a CMH is contained in the ORNL engineering drawings listed in the appropriate sections. The components are fabricated and inspected from these drawings in accordance with a Quality Assurance Plan and a Manufacturing Plan. The material in this report describes the acceptance and performance tests of CMH subassemblies used ty the Tower Shielding Facility (TSF) staff but it can also be used by personnel fabricating the components. This information which was developed and used before the advent of the formalized QA Program and Manufacturing Plans evolved during the fabrication and testing of the first five CMHs

Creep property testing of energy power plant component material

International Nuclear Information System (INIS)

Nitiswati, Sri; Histori; Triyadi, Ari; Haryanto, Mudi

1999-01-01

Creep testing of SA213 T12 boiler piping material from fossil plant, Suralaya has been done. The aim of the testing is to know the creep behaviour of SA213 T12 boiler piping material which has been used more than 10 yeas, what is the material still followed ideal creep curve (there are primary stage, secondary stage, and tertiary stage). This possibility could happened because the material which has been used more than 10 years usually will be through ageing process because corrosion. The testing was conducted in 520 0C, with variety load between 4% until 50% maximum allowable load based on strength of the material in 520 0C

Philips Electronics synchronizes its supply chain to end the bullwhip effect

NARCIS (Netherlands)

Kok, de A.G.; Janssen, F.B.S.L.P.; van Doremalen, J.B.M.; Wachem, van E.; Clerkx, M.J.R.; Peeters, W.

2005-01-01

Demand variability increases as one moves up a supply chain. The demand for finished products is less variable than for subassemblies, which is less variable than for individual components. This phenomenon is known as the bullwhip or Forrester effect. It increases inventory unnecessarily and makes

Material Selection for an Ultra High Strength Steel Component Based on the Failure Criteria of CrachFEM

International Nuclear Information System (INIS)

Kessler, L.; Beier, Th.; Werner, H.; Horstkott, D.; Dell, H.; Gese, H.

2005-01-01

An increasing use of combining more than one process step is noticed for coupling crash simulations with the results of forming operations -- mostly by inheriting the forming history like plastic strain and material hardening. Introducing a continuous failure model allows a further benefit of these coupling processes; it sometimes can even be the most attractive result of such a work. In this paper the algorithm CrachFEM for fracture prediction has been used to generate more benefit of the successive forming and crash simulations -- especially for ultra high strength steels. The choice and selection of the material grade in combination with the component design can therefore be done far before the prototyping might show an unsuccessful crash result; and in an industrial applicable manner

Identification and Assessment of Material Models for Age-Related Degradation of Structures and Passive Components in Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Nie,J.; Braverman, J.; Hofmayer, C.; Kim, M. K.; Choi, I-K.

2009-04-27

describes the research effort performed by BNL for the Year 2 scope of work. This research focused on methods that could be used to represent the long-term behavior of materials used at NPPs. To achieve this BNL reviewed time-dependent models which can approximate the degradation effects of the key materials used in the construction of structures and passive components determined to be of interest in the Year 1 effort. The intent was to review the degradation models that would cover the most common time-dependent changes in material properties for concrete and steel components.

Lifetime analysis of fusion-reactor components

International Nuclear Information System (INIS)

Mattas, R.F.

1983-01-01

A one-dimensional computer code has been developed to examine the lifetime of first-wall and impurity-control components. The code incorporates the operating and design parameters, the material characteristics, and the appropriate failure criteria for the individual components. The major emphasis of the modelling effort has been to calculate the temperature-stress-strain-radiation effects history of a component so that the synergystic effects between sputtering erosion, swelling, creep, fatigue, and crack growth can be examined. The general forms of the property equations are the same for all materials in order to provide the greatest flexibility for materials selection in the code. The code is capable of determining the behavior of a plate, composed of either a single or dual material structure, that is either totally constrained or constrained from bending but not from expansion. The code has been utilized to analyze the first walls for FED/INTOR and DEMO

Creep fatigue design of FBR components

International Nuclear Information System (INIS)

Bhoje, S.B.; Chellapandi, P.

1997-01-01

This paper deals with the characteristic features of Fast Breeder Reactor (FBR) with reference to creep fatigue, current creep fatigue design approach in compliance with RCCMR (1987) design code, material data, effects of weldments and neutron irradiation, material constitutive models employed, structural analysis and further R and D required for achieving maturity in creep fatigue design of FBR components. For the analysis reported in this paper, material constitutive models developed based on ORNIb (Oak Ridge National Laboratory) and Chaboche viscoplastic theories are employed to demonstrate the potential of FBR components for higher plant temperatures and/or longer life. The results are presented for the studies carried out towards life prediction of Prototype Fast Breeder Reactor (PFBR) components. (author). 24 refs, 8 figs, 5 tabs

Philosophy of integrity assessment of engineering components

International Nuclear Information System (INIS)

Chaudhuri, Satyabrata

2008-01-01

Integrity assessment of engineering components in power plants and process industries has attracted global attention from the viewpoint of safety and economy for their optimum utilization. This paper describes some aspects of philosophy of component integrity such as life assessment technology, materials used and the factors limiting the serviceability of the components operating at high temperatures and pressures. Numerous investigations have been carried out all over the world to study changes in microstructure and material property due to prolonged service of the components to decide their further serviceability. This paper includes case studies on integrity assessment of service-exposed components carried out in our laboratory as well

26 CFR 1.460-1 - Long-term contracts.

Science.gov (United States)

2010-04-01

... the manufacture of personal property is a manufacturing contract. In contrast, a contract for the... performance of engineering and design services, and the production of components and subassemblies that are..., enters into a single long-term contract to design and manufacture a satellite and to develop computer...

Geochemical and Sr-Nd-Pb-Li isotopic characteristics of volcanic rocks from the Okinawa Trough: Implications for the influence of subduction components and the contamination of crustal materials

Science.gov (United States)

Guo, Kun; Zhai, Shikui; Yu, Zenghui; Wang, Shujie; Zhang, Xia; Wang, Xiaoyuan

2018-04-01

The Okinawa Trough is an infant back-arc basin developed along the Ryukyu arc. This paper provides new major and trace element and Sr-Nd-Pb-Li isotope data of volcanic rocks in the Okinawa Trough and combines the published geochemical data to discuss the composition of magma source, the influence of subduction component, and the contamination of crustal materials, and calculate the contribution between subduction sediment and altered oceanic crust in the subduction component. The results showed that there are 97% DM and 3% EMI component in the mantle source in middle trough (MS), which have been influenced by subduction sediment. The Li-Nd isotopes indicate that the contribution of subduction sediment and altered oceanic crust in subduction component are 4 and 96%, respectively. The intermediate-acidic rocks suffer from contamination of continental crust material in shallow magma chamber during fractional crystallization. The acidic rocks in south trough have experienced more contamination of crustal material than those from the middle and north trough segments.

Explosive Components Facility

Data.gov (United States)

Federal Laboratory Consortium — The 98,000 square foot Explosive Components Facility (ECF) is a state-of-the-art facility that provides a full-range of chemical, material, and performance analysis...

Magnetic fusion energy plasma interactive and high heat flux components. Volume II. Technical assessment of the critical issues and problem areas in high heat flux materials and component development

International Nuclear Information System (INIS)

Abdou, M.A.; Boyd, R.D.; Easor, J.R.

1984-06-01

A technical assessment of the critical issues and problem areas for high heat flux materials and components (HHFMC) in magnetic fusion devices shows these problems to be of critical importance for the successful operation of near-term fusion experiments and for the feasibility and attractiveness of long-term fusion reactors. A number of subgroups were formed to assess the critical HHFMC issues along the following major lines: (1) source conditions, (2) systems integration, (3) materials and processes, (4) thermal hydraulics, (5) thermomechanical response, (6) electromagnetic response, (7) instrumentation and control, and (8) test facilities. The details of the technical assessment are presented in eight chapters. The primary technical issues and needs for each area are highlighted

Magnetic fusion energy plasma interactive and high heat flux components. Volume II. Technical assessment of the critical issues and problem areas in high heat flux materials and component development

Energy Technology Data Exchange (ETDEWEB)

Abdou, M.A.; Boyd, R.D.; Easor, J.R.; Gauster, W.B.; Gordon, J.D.; Mattas, R.F.; Morgan, G.D.; Ulrickson, M.A,; Watson, R.D.; Wolfer, W.G,

1984-06-01

A technical assessment of the critical issues and problem areas for high heat flux materials and components (HHFMC) in magnetic fusion devices shows these problems to be of critical importance for the successful operation of near-term fusion experiments and for the feasibility and attractiveness of long-term fusion reactors. A number of subgroups were formed to assess the critical HHFMC issues along the following major lines: (1) source conditions, (2) systems integration, (3) materials and processes, (4) thermal hydraulics, (5) thermomechanical response, (6) electromagnetic response, (7) instrumentation and control, and (8) test facilities. The details of the technical assessment are presented in eight chapters. The primary technical issues and needs for each area are highlighted.

Destructive Examination of Shipping Package 9975-02019

Energy Technology Data Exchange (ETDEWEB)

Daugherty, W. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-06-13

Destructive and non-destructive examinations have been performed on the components of shipping package 9975-02019 as part of a comprehensive SRS surveillance program for plutonium material stored in the K-Area Complex (KAC). During the field surveillance inspection of this package in KAC, two non-conforming conditions were noted: the axial gap of 1.577 inch exceeded the 1 inch maximum criterion, and two areas of dried glue residue were noted on the upper fiberboard subassembly. This package was subsequently transferred to SRNL for more detailed inspection and destructive examination. In addition to the conditions noted in KAC, the following conditions were noted: - Numerous small spots of corrosion were observed along the bottom edge of the drum. - In addition to the smeared glue residue on the upper fiberboard subassembly, there was also a small dark stain. - Mold was present on the side and bottom of the lower fiberboard subassembly. Dark stains from elevated moisture content were also present in these areas. - A dark spot with possible light corrosion was observed on the primary containment vessel flange, and corresponding rub marks were observed on the secondary containment vessel ID. - The fiberboard thermal conductivity in the radial orientation was above the specified range. When the test was repeated with slightly lower moisture content, the result was acceptable. The moisture content for both tests was within a range typical of other packages in storage. The observed conditions must be fully evaluated by KAC to ensure the safety function of the package is being maintained. Several factors can contribute to the concentration of moisture in the fiberboard, including higher than average initial moisture content, higher internal temperature (due to internal heat load and placement within the array of packages), and the creation of additional moisture as the fiberboard begins to degrade.

High-Temperature Corrosion Study for the RPP Low Activity Waste Melter

International Nuclear Information System (INIS)

Marshall, K.M.

2003-01-01

The River Protection Program (RPP) low activity waste (LAW) melter design incorporates a series of bubblers used to increase convection in the molten glass. Through runs of a pilot melter at Duratek, Inc. in Columbia, Maryland, the bubblers have been identified as the major component limiting LAW melter availability, requiring frequent replacement due to corrosive degradation, primarily at the melt line. Laboratory experiments were performed to evaluate the performance of several alloys and coatings in simulated RPP low activity waste melter vapor space and molten glass environments. The performance of the alloys and coatings was studied in order to advance our understanding of how these materials react at the melt/air interface inside the melter. The ultimate goal was to identify a material with superior performance compared to that of Inconel 693, and to deliver a bubbler sub-assembly made of that material to the RPP LAW melter pilot facility for further testing

Increased component safety through improved methods for residual stress analysis. Subprojects. Consideration of the elastic-plastic material properties (Phase 2). Final report

International Nuclear Information System (INIS)

Mirbach, David von

2015-01-01

Residual stresses in mechanical components can result in both detrimental but also beneficial effects on the strength and lifetime of the components. The most detailed knowledge of the residual stress state is of advantage or a pre-requisite for the assessment of the component performance. Two commonly used methods for determination of residual stresses are the hole drilling method and the ring core method which can be regarded to the mechanical methods. In the context of reactor safety research of the German Federal Ministry of Economic and Energy (BMWi) two fundamental and interacting weak points of the hole drilling method as well as of the ring core method, respectively, in order to determine residual stresses are going to be investigated. As a consequence reliability of the methods will be improved in this joint research project. On the one hand there are effects of geometrical boundary conditions of the components and on the other hand there is the influence of plasticity due to notch effects both affecting the released strain field after removing material and after all the calculated residual stresses. The first issue mentioned above is under the responsibility of the Institute of Materials Engineering (Kassel University) and the last one is investigated by materials testing institute university Stuttgart. As a consequence of a successful project the knowledge base will be considerably improved resulting in benefits for various engineering fields. Especially the quantitative consideration of real residual stress states for optimized component designs will be possible and after all the consequences of residual stresses on safety of components which are used in nuclear facilities can be evaluated. In this second experimental research chapter (phase 2) the findings of the first numerical and theoretical research chapter (phase 1) where proofed. The developed differential calculation method with the method of adaptive calibration functions were compared with the

INVESTIGATION OF THE PRESENCE OF DRUGSTORE BEETLES WITHIN CELOTEX ASSEMBLIES IN RADIOACTIVE MATERIAL PACKAGINGS

Energy Technology Data Exchange (ETDEWEB)

Loftin, B; Glenn Abramczyk, G

2008-06-04

During normal operations at the Department of Energy's Hanford Site in Hanford, WA, drugstore beetles, (Stegobium paniceum (L.) Coleoptera: Anobiidae), were found within the fiberboard subassemblies of two 9975 Shipping Packages. Initial indications were that the beetles were feeding on the Celotex{trademark} assemblies within the package. Celotex{trademark} fiberboard is used in numerous radioactive material packages serving as both a thermal insulator and an impact absorber for both normal conditions of transport and hypothetical accident conditions. The Department of Energy's Packaging Certification Program (EM-63) directed a thorough investigation to determine if the drugstore beetles were causing damage that would be detrimental to the safety performance of the Celotex{trademark}. The Savannah River National Laboratory is conducting the investigation with entomological expertise provided by Clemson University. The two empty 9975 shipping packages were transferred to the Savannah River National Laboratory in the fall of 2007. This paper will provide details and results of the ongoing investigation.

30 CFR 27.41 - Test to determine resistance to moisture.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Test to determine resistance to moisture. 27.41... determine resistance to moisture. Components, subassemblies, or assemblies, the normal functioning of which might be affected by moisture, shall be tested in atmospheres of high relative humidity (80 percent or...

PRINCIPLES OF RE-ENGINEERING METHODOLOGY FOR TECHNOLOGICAL PROCESS IN PROCESSING OF RAW MATERIAL COMPONENTS WHILE PRODUCING CEMENT AND SILICATE PRODUCTS

Directory of Open Access Journals (Sweden)

I. A. Busel

2014-01-01

Full Text Available Grinding process is characterized by high energy consumption and low productivity. Nowadays efficiency of the ball mills applied for grinding is rather low. Only 3-6 % of the supplied power energy is used for material grinding. The rest part of the energy disappears in the form of heat, vibration and noise. So matter concerning reduction of energy consumption is of great importance.Improvement of efficiency and quality of technological process in grinding of raw material components while producing construction materials is considered as one of priority-oriented targets of power- and resource saving in construction industry with the purpose to reduce energy consumption for grinding. Grinding efficiency at operating enterprises is reasonable to improve by modernization of the equipment and existing technological, management and other processes which are related to grinding of mineral raw material. In order to reduce grinding power consumption it is necessary to carry out a complex re-engineering of technological process in grinding of various materials which is based on usage of new modifications of grinding bodies, physical and chemical grinding aids, modern information technologies and industrial automation equipment. Application of modern information technologies and industrial automation equipment makes it possible to execute the grinding process with maximum achievable productivity for existing capacity due to automatic control and consideration of continuous changes in technological parameters. In addition to this such approach gives an opportunity to control processes in real time by immediate adjustments of technological equipment operational modes.The paper considers an approach to the development of re-engineering methodology for technological process in grinding of raw material components while producing construction materials. The present state of technological grinding process is presented in the paper. The paper points out the

3D-printed optical active components

Science.gov (United States)

Suresh Nair, S.; Nuding, J.; Heinrich, A.

2018-02-01

Additive Manufacturing (AM) has the potential to become a powerful tool in the realization of complex optical components. The primary advantage that meets the eye, is that fabrication of geometrically complicated optical structures is made easier in AM as compared to the conventional fabrication methods (using molds for instance). But this is not the only degree of freedom that AM has to offer. With the multitude of materials suitable for AM in the market, it is possible to introduce functionality into the components one step before fabrication: by altering the raw material. A passive example would be to use materials with varying properties together, in a single manufacturing step, constructing samples with localized refractive indices for instance. An active approach is to blend in materials with distinct properties into the photopolymer resin and manufacturing with this composite material. Our research is currently focused in this direction, with the desired optical property to be introduced being Photoluminescence. Formation of nanocomposite mixtures to produce samples is the current approach. With this endeavor, new sensor systems can be realized, which may be used to measure the absorption spectra of biological samples. Thereby the sample compartment, the optics and the spectral light source (different quantum dots) are 3D-printed in one run. This component can be individually adapted to the biological sample with respect to wavelength, optical and mechanical properties. Here we would like to present our work on the additive manufacturing of an active optical component. Based on the stereolithography method, a monolithic optical component was 3D-printed, showing light emission at different defined wavelengths due to UV excited quantum dots inside the 3D-printed optics.

The relationship between material fracture resistance and the kinetics of fracture in steel components

International Nuclear Information System (INIS)

Irvine, W.H.

1978-01-01

The conditions necessary for the onset of fast brittle fracture are reasonably well understood. However with increasing material ductility at normal engineering stress levels the effects of structure size and type of loading become more important and make the understanding of the behaviour of large structures and laboratory test pieces and their inter-relation, more difficult.By using Berry's concept of a fracture locus, it is shown that the crack size - stress level - material fracture resistance relationship, as typified for instance by the Griffith-Irwin formulae, is necessary and sufficient for defining the point at which fast brittle fracture occurs, but that in the case of fast ductile fracture it is not sufficient by itself and must be supplemented by a description of the unloading path of the structural system. Although the demarcation line between these two types of behaviour is seen to be dependent on stress level it can nevertheless provide a definition of brittle and ductile fracture in engineering structures. Berry's use of the Griffith equation to describe the separation of the crack tip material limits any practical use of his locus equation to stress levels that are low by present day engineering standards. Consideration is given to the use of relationships describing crack tip failure which are more appropriate for the ductilities and stress levels of current engineering interest. These equations explicitly involve the size of the crack tip perturbation and therefore allow a direct check to be made on validity. Examples are given of the application of these methods to describe fractures which have occurred in structural components. (author)

PIE technology on mechanical tests for HTTR core component and structural materials developed at Research Hot Laboratory

International Nuclear Information System (INIS)

Kizaki, Minoru; Honda, Junichi; Usami, Kouji; Ouchi, Asao; Oeda, Etsuro; Matsumoto, Seiichiro

2001-02-01

The high temperature engineering test reactor (HTTR) with the target operation temperature of 950degC established the first criticality on November, 1998 based on a large amount of R and D results on fuel and materials. In such R and D works, the development of reactor materials are one of the key issues from the view point of reactor environments such as extremely high temperature, neutron irradiation and so on for the HTTR. The Research Hot Laboratory (RHL) had carried out much kind of post irradiation examinations (PIEs) on core component and pressure vessel materials for during more than a quarter century. And obtained data played an important role in development, characterization and licensing of those materials for the HTTR. This paper describes the PIE technology developed at RHL and typical results on mechanical tests such as elevated temperature tensile and creep rupture tests for Hasteloy-X, Incolloy 800H and so on, and Charpy impact, J IC fracture toughness, K Id fracture toughness and small punch tests for normalized and tempered 2 1/4Cr-1Mo steel from historical view. In addition, an electrochemical test technique established for investigating the irradiation embrittlement mechanism on 2 1/4Cr-1Mo steel is also mentioned. (author)

Modern electronic materials

CERN Document Server

Watkins, John B

2013-01-01

Modern Electronic Materials focuses on the development of electronic components. The book first discusses the history of electronic components, including early developments up to 1900, developments up to World War II, post-war developments, and a comparison of present microelectric techniques. The text takes a look at resistive materials. Topics include resistor requirements, basic properties, evaporated film resistors, thick film resistors, and special resistors. The text examines dielectric materials. Considerations include basic properties, evaporated dielectric materials, ceramic dielectri

Theoretical and experimental investigations concerning the problem of quasi-static crack propagation in two-component materials subject to residual stresses

International Nuclear Information System (INIS)

Braun, H.P.

1979-01-01

With the aim of obtaining microstructural information of multi-component materials fracture-mechanical calculations on continuum-mechanical models of fiber composites were performed. There were ideal sections of material permitting the formulation of suitable mixed boundary value problems of static thermoelasticity whose solutions could be obtained by means of appropriate numerical methods from continuum mechanics. As model loads exclusively thermally induced residual stresses were assumed, being of special interest because of the thermomechanically inhomogeneous structure of composite materials on one hand and having got decisive significance for a number of important areas of application as e.g. aero-space industry, reactor technology and chemical apparatus engineering on the other. The results evaluated numerically are exemplarily examined by means of photoelasticity. (orig./IHOE) [de

In-service materials testing of selected components of unit 1 and 2 of V-1 nuclear power plant

International Nuclear Information System (INIS)

Cintula, J.

1982-01-01

The task of in-service nondestructive testing of nuclear installations is to confirm that the state of base material and welded joints has not changed owing to mechanical, thermal or radiation stress. Under the regulations of safe operation the first in-service inspection of all components of a WWER 440 reactor must be carried out after 15,000 to 2O,00O operating hours at the latest. Further in-service inspections are repeated after 30,000 hours (pressure vessels) and 40,000 hours (the main steam piping and the feedwater piping). Proceeding from experience gained so far, intervals are suggested for in-service checks of the other components of the V-1 nuclear power plant. Also briefly described are the main nondestructive methods used for such checks at this power plant. (Z.M.)

Metallic materials for heat exchanger components and highly stressed internal of HTR reactors for nuclear process heat generation

International Nuclear Information System (INIS)

1982-01-01

The programme was aimed at the development and improvement of materials for the high-temperature heat exchanger components of a process steam HTR. The materials must have high resistance to corrosion, i.e. carburisation and internal oxidation, and high long-term toughness over a wide range of temperatures. They must also meet the requirements set in the nuclear licensing procedure, i.e. resistance to cyclic stress and irradiation, non-destructive testing, etc. Initially, it was only intended to improve and qualify commercial alloys. Later on an alloy development programme was initiated in which new, non-commercial alloys were produced and modified for use in a nuclear process heat facility. Separate abstracts were prepared for 19 pays of this volume. (orig./IHOE) [de

Structural integrity testing of glass-ceramic/molybdenum vacuum tube frames

International Nuclear Information System (INIS)

Spears, R.K.

1980-01-01

In this study, vacuum tube subassemblies made of glass-ceramic insulators sealed to inner and outer molybdenum frames were loaded in compression to failure with a tensile test machine. Several factors were varied in processing these subassemblies. These factors included etching and nonetching of molybdenum piece parts, annealing and nonannealing of subassemblies, and vapor and non-vapor honing of insulators after sealing. After failure, the subassemblies were examined for fracture patterns. In most cases, fracture started at points near the lower portion of the inner sleeve-insulator interface. More load was carried by subassemblies having molybdenum piece parts that were acid etched. No difference appeared between the strength of subassemblies having annealed and nonannealed glass-ceramic insulators. Parts with vapor-honed insulators failed at substantially lower loads

ARCHER Project: Progress on Material and component activities for the Advanced High Temperature Reactor

International Nuclear Information System (INIS)

Buckthorpe, D.E.

2014-01-01

The ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R&D) integrated project is a four year project which was started in 2011 as part of the European Commission 7th Framework Programme (FP7) to perform High Temperature Reactor technology R&D in support of reactor demonstration. The project consortium encompasses conventional and Nuclear Industry, Utilities, Technical Support Organizations, Research & Development Organizations and Academia. The activities involved contribute to the Generation IV (GIF) International Forum and collaborate with related projects in the US, China, Japan, and the Republic of Korea in cooperation with IAEA and ISTC. This paper addresses the progress of the work on ARCHER materials and component activities since the start of the project and underlines some of the main conclusions reached. (author)

Modified cleaning method for biomineralized components

Science.gov (United States)

Tsutsui, Hideto; Jordan, Richard W.

2018-02-01

The extraction and concentration of biomineralized components from sediment or living materials is time consuming and laborious and often involves steps that remove either the calcareous or siliceous part, in addition to organic matter. However, a relatively quick and easy method using a commercial cleaning fluid for kitchen drains, sometimes combined with a kerosene soaking step, can produce remarkable results. In this study, the method is applied to sediments and living materials bearing calcareous (e.g., coccoliths, foraminiferal tests, holothurian ossicles, ichthyoliths, and fish otoliths) and siliceous (e.g., diatom valves, silicoflagellate skeletons, and sponge spicules) components. The method preserves both components in the same sample, without etching or partial dissolution, but is not applicable to unmineralized components such as dinoflagellate thecae, tintinnid loricae, pollen, or plant fragments.

Explosion bonding of dissimilar materials for fabricating APS front end components: Analysis of metallurgical and mechanical properties and UHV applications

International Nuclear Information System (INIS)

Li, Yuheng; Shu, Deming; Kuzay, T.M.

1994-01-01

The front end beamline section contains photon shutters and fixed masks. These components are made of OFHC copper and GlidCOP AL-15. Stainless steels (304 or 316) are also used for connecting photon shutters and fixed masks to other components that operate in the ultrahigh vacuum system. All these dissimilar materials need to be joined together. However, bonding these dissimilar materials is very difficult because of their different mechanical and thermal properties and incompatible metallurgical properties. Explosion bonding is a bonding method in which the controlled energy of a detonating explosive is used to create a metallurgical bond between two or more similar or dissimilar materials. No intermediate filler metal, for example, a brazing compound or soldering alloy, is needed to promote bonding, and no external heat need be applied. A study of the metallurgical and mechanical properties and YGV applications of GlidCop AL-15, OFHC copper, and 304 stainless steel explosion-bonded joints has been done. This report contains five parts: an ultrasonic examination of explosion-bonded joints and a standard setup; mechanical-property and thermal-cycle tests of GlidCop AL-15/304 stainless steel explosion-bonded joints; leak tests of a GlidCop AL-15/304 stainless steel explosion-bonded interfaces for UHV application; metallurgical examination of explosion-bonded interfaces and failure analysis, and discussion and conclusion

The potential for reducing the radiological consequences of reactor decommissioning through selection of construction materials for activated components

International Nuclear Information System (INIS)

Woollam, P.B.

1984-08-01

This report considers whether it may be possible to reduce the radiological consequences of reactor decommissioning by careful attention to the specification of the elemental concentration of materials used in the reactor's construction. In particular, consideration is given to the potential for reduction of the concentration of elements known to activate to long lived daughter isotopes. Two particular areas are addressed, both applied to Sizewell 'B' PWR. The first is the choice of raw materials for the construction of the concrete bioshield to minimise future waste arisings. The second is the specification of some trace element concentrations in the steel pressure vessel and reactor internal structures to minimise personnel exposure at decommissioning time. The report presents extensive analyses of many of the candidate raw materials for Sizewell 'B' concrete, including PFA, and derives the radiological consequences for the eventual disposal of these materials to a hypothetical municipal land fill waste site. Data are also presented on the concentrations of important elements activating to gamma emitting daughters in type 304 stainless steels, leading to an assessment of likely dose equivalent rates at decommissioning time from the pressure vessel and from the internal components. (author)

The computer code EURDYN - 1 M (release 1) for transient dynamic fluid-structure interaction. Pt.1: governing equations and finite element modelling

International Nuclear Information System (INIS)

Donea, J.; Fasoli-Stella, P.; Giuliani, S.; Halleux, J.P.; Jones, A.V.

1980-01-01

This report describes the governing equations and the finite element modelling used in the computer code EURDYN - 1 M. The code is a non-linear transient dynamic program for the analysis of coupled fluid-structure systems; It is designed for safety studies on LMFBR components (primary containment and fuel subassemblies)

Artificial heart system thermal converter and blood pump component research and development

International Nuclear Information System (INIS)

Pouchot, W.D.; Bifano, N.J.; Hanson, J.P.

1975-01-01

A bench model version of a nuclear-powered artificial heart system to be used as a replacement for the natural heart was constructed and tested as a part of a broader U. S. ERDA program. The objective of the broader program has been to develop a prototype of a fully implantable nuclear-powered total artificial heart system powered by the thermal energy of plutonium-238 and having minimum weight and volume and a minimum life of ten years. As a forward step in this broader program, component research and development has been carried out directed towards a fully implantable and advanced version of the bench model (IVBM). Some of the results of the component research and development effort on a Stirling engine, blood pump drive mechanisms, and coupling mechanisms are presented. The Stirling-mechanical system under development is shown. There are three major subassemblies: the thermal converter, the coupling mechanism, and the blood pump drive mechanism. The thermal converter uses a Stirling cycle to convert the heat of the plutonium-238 fueled heat source to a rotary shaft power output. The coupling mechanism changes the orientation of the output shaft by 90 degrees and transmits the pumping power by wire-wound core flexible shafting to the pumping mechanism. The coupling mechanism also provides routing of the coolant lines which carry the cycle waste heat from the thermal converter to the blood pump. The change in orientation of the thermal converter output shaft is for convenience in implanting in a calf. This orientation of thermal converter to blood pump seemed to give the best overall system fit in a calf based on fit trials with wooden models in a calf cadaver

Bioactive Materials in Endodontics: An Evolving Component of Clinical Dentistry.

Science.gov (United States)

Mohapatra, Satyajit; Patro, Swadheena; Mishra, Sumita

2016-06-01

Achieving biocompatibility in a material requires an interdisciplinary approach that involves a sound knowledge of materials science, bioengineering, and biotechnology. The host microbial-material response is also critical. Endodontic treatment is a delicate procedure that must be planned and executed properly. Despite major advances in endodontic therapy in recent decades, clinicians are confronted with a complex root canal anatomy and a wide selection of endodontic filling materials that, in turn, may not be well tolerated by the periapical tissues and may evoke an immune reaction. This article discusses published reports of various bioactive materials that are used in endodontic therapy, including calcium hydroxide, mineral trioxide aggregate, a bioactive dentin substrate, calcium phosphate ceramics, and calcium phosphate cements.

Energy Conversion Alternatives Study (ECAS), General Electric Phase 1. Volume 3: Energy conversion subsystems and components. Part 1: Bottoming cycles and materials of construction

Science.gov (United States)

Shah, R. P.; Solomon, H. D.

1976-01-01

Energy conversion subsystems and components were evaluated in terms of advanced energy conversion systems. Results of the bottoming cycles and materials of construction studies are presented and discussed.

Calcination and solid state reaction of ceramic-forming components to provide single-phase superconducting materials having fine particle size

Science.gov (United States)

Balachandran, Uthamalingam; Poeppel, Roger B.; Emerson, James E.; Johnson, Stanley A.

1992-01-01

An improved method for the preparation of single phase, fine grained ceramic materials from precursor powder mixtures where at least one of the components of the mixture is an alkali earth carbonate. The process consists of heating the precursor powders in a partial vacuum under flowing oxygen and under conditions where the partial pressure of CO.sub.2 evolved during the calcination is kept to a very low level relative to the oxygen. The process has been found particularly suitable for the preparation of high temperature copper oxide superconducting materials such as YBa.sub.2 Cu.sub.3 O.sub.x "123" and YBa.sub.2 Cu.sub.4 O.sub.8 "124".

Pulsed laser deposition of chalcogenide sulfides from multi- and single-component targets: the non-stoichiometric material transfer

DEFF Research Database (Denmark)

Schou, Jørgen; Ganskukh, Mungunshagai; Ettlinger, Rebecca Bolt

2018-01-01

The mass transfer from target to films is incongruent for chalcogenide sulfides in contrast to the expectations of pulsed laser deposition (PLD) as a stoichiometric film growth process. Films produced from a CZTS (Cu2ZnSnS4) multi-component target have no Cu below a fluence threshold of 0.2 J/cm2......, and the Cu content is also very low at low fluence from a single-component target. Above this threshold, the Cu content in the films increases almost linearly up to a value above the stoichiometric value, while the ratio of the concentration of the other metals Zn to Sn (Zn/Sn) remains constant. Films...... of a similar material CTS (Cu2SnS3) have been produced by PLD from a CTS target and exhibits a similar trend in the same fluence region. The results are discussed on the basis of solid-state data and the existing data from the literature....

Flow induced vibration studies for LMFBR in Japan: Past and recent studies of FIV for JOYO and MONJU

Energy Technology Data Exchange (ETDEWEB)

Sato, K [Sodium Engineering Division, O-arai Engineering Centre, Power Reactor and Nuclear Fuel, Development Corporation, Narita-cho, O-arai Machi, Ibaraki-ken (Japan)

1977-12-01

This paper presents the past and recent studies of flow induced vibration of the reactor components for the experimental fast breeder reactor JOYO and the prototype fast breeder reactor MONJU, in which many suggestive results for the higher flow velocity systems in a future reactor are contained. The fuel subassembly is the most important from the view point of the vibration. Thus, the studies were carried out with a mock-up subassembly for JOYO. In this experiment, statistical analysis results of the vibration characteristics of single core subassembly and the effects of external forced vibration, flow disturbance and fuel pin bundle vibration were reported. The further more detailed investigations are now being performed for MONJU. In addition to the above studies, the vibration failure of a sodium valve is reported. The valve is a 8-inch stop valve in SODIUM FLOW AND HEAT TRANSFER TEST LOOP at O-arai Engineering Center. The failure occurred in 1969 during the performance test of the mechanical pump, and this resulted in a small sodium leak. The cause of the failure was found to be the vibration fatigue of the metal bellows. (author)

Mechanical thermal and electric measurements on materials and components of the main coils of the Milan superconducting cyclotron

International Nuclear Information System (INIS)

Acerbi, E.; Rossi, L.

1988-01-01

The coils of the Milan Superconducting Cyclotron are the largest superconducting devices built up to now in Italy and constitute the first superconducting magnet for accelerator in Europe. Because of the large stored energy (more than 40 MJ), of the high stresses and of of the need of reliability, a lot of measurements were carried out as well on materials used for the coils, both on superconducting cable and structural materials, as on the main components of the coils and on two double pancakes prototypes (wound with full copper cable). In this paper the results on these measurements are reported and the results of tests on the prototypes are discussed. The aim is to provide an easy source of data for superconducting coils useful to verify calculations or to improve the performances

Lessons Learned in Risk Management on NCSX

International Nuclear Information System (INIS)

Neilson, G.H.; Gruber, C.O.; Harris, Jeffrey H.; Rej, D.J.; Simmons, R.T.; Strykowsky, R.L.

2010-01-01

The National Compact Stellarator Experiment (NCSX) was designed to test physics principles of an innovative stellarator design developed by Princeton Plasma Physics Laboratory and Oak Ridge National Laboratory. Construction of some of the major components and subassemblies was completed, but the estimated cost and schedule for completing the project grew as the technical requirements and risks became better understood, leading to its cancellation in 2008. The project's risks stemmed from its technical challenges, primarily the complex component geometries and tight tolerances that were required. The initial baseline, which was established in 2004, was supported by a risk management plan and risk-based contingencies, both of which proved to be inadequate. Technical successes were achieved in the construction of challenging components and subassemblies, but cost and schedule growth was experienced. As part of an effort to improve project performance, a new risk management program was devised and implemented in 2007-2008. It led to a better understanding of project risks, a sounder basis for contingency estimates, and improved management tools. Although the risks were ultimately unacceptable to the sponsor, valuable lessons in risk management were learned through the experiences with the NCSX project.

Lessons Learned in Risk Management on NCSX

International Nuclear Information System (INIS)

Neilson, G.H.; Gruber, C.O.; Harris, J.H.; Rej, D.J.; Simmons, R.T.; Strykowsky, R.L.

2009-01-01

The National Compact Stellarator Experiment (NCSX) was designed to test physics principles of an innovative stellarator design developed by the Princeton Plasma Physics Laboratory and Oak Ridge National Laboratory. Construction of some of the major components and sub-assemblies was completed, but the estimated cost and schedule for completing the project grew as the technical requirements and risks became better understood, leading to its cancellation in 2008. The project's risks stemmed from its technical challenges, primarily the complex component geometries and tight tolerances that were required. The initial baseline, established in 2004, was supported by a risk management plan and risk-based contingencies, both of which proved to be inadequate. Technical successes were achieved in the construction of challenging components and subassemblies, but cost and schedule growth was experienced. As part of an effort to improve project performance, a new risk management program was devised and implemented in 2007-08. It led to a better understanding of project risks, a sounder basis for contingency estimates, and improved management tools. Although the risks ultimately were unacceptable to the sponsor, valuable lessons in risk management were learned through the experiences with the NCSX project

Laser materials processing of complex components. From reverse engineering via automated beam path generation to short process development cycles.

Science.gov (United States)

Görgl, R.; Brandstätter, E.

2016-03-01

The article presents an overview of what is possible nowadays in the field of laser materials processing. The state of the art in the complete process chain is shown, starting with the generation of a specific components CAD data and continuing with the automated motion path generation for the laser head carried by a CNC or robot system. Application examples from laser welding, laser cladding and additive laser manufacturing are given.

Laser materials processing of complex components: from reverse engineering via automated beam path generation to short process development cycles

Science.gov (United States)

Görgl, Richard; Brandstätter, Elmar

2017-01-01

The article presents an overview of what is possible nowadays in the field of laser materials processing. The state of the art in the complete process chain is shown, starting with the generation of a specific components CAD data and continuing with the automated motion path generation for the laser head carried by a CNC or robot system. Application examples from laser cladding and laser-based additive manufacturing are given.

Managing traceability information in manufacture

NARCIS (Netherlands)

Jansen-Vullers, M.H.; Dorp, van C.A.; Beulens, A.J.M.

2003-01-01

In this paper, an approach to design information systems for traceability is proposed. The paper applies gozinto graph modelling for traceability of the goods flow. A gozinto graph represents a graphical listing of raw materials, parts, intermediates and subassemblies, which a process transforms

Managing traceability information manufacture

NARCIS (Netherlands)

Jansen-Vullers, M.H.; van Dorp, C.A.; Beulens, A.J.M.

2003-01-01

In this paper, an approach to design information systems for traceability is proposed. The paper applies gozinto graph modelling for traceability of the goods flow. A gozinto graph represents a graphical listing of raw materials, parts, intermediates and subassemblies, which a process transforms

Liquid metal cooled fast breeder nuclear reactors

International Nuclear Information System (INIS)

Thatcher, G.; Mitchell, A.J.

1981-01-01

Fuel sub-assemblies for liquid metal-cooled fast breeder reactors are described which each incorporate a fluid flow control valve for regulating the rate of flow through the sub-assembly. These small electro-magnetic valves seek to maintain the outlet coolant temperature of at least some of the breeder sub-assemblies substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (U.K.)

Beam-Material Interaction

CERN Document Server

Mokhov, N.V.

2016-01-01

Th is paper is motivated by the growing importance of better understanding of the phenomena and consequences of high- intensity energetic particle beam interactions with accelerator, generic target , and detector components. It reviews the principal physical processes of fast-particle interactions with matter, effects in materials under irradiation, materials response, related to component lifetime and performance, simulation techniques, and methods of mitigating the impact of radiation on the components and envir onment in challenging current and future application

Beam-Material Interaction

Energy Technology Data Exchange (ETDEWEB)

Mokhov, N. V. [Fermilab; Cerutti, F. [CERN

2016-01-01

Th is paper is motivated by the growing importance of better understanding of the phenomena and consequences of high-intensity energetic particle beam interactions with accelerator, generic target, and detector components. It reviews the principal physical processes of fast-particle interactions with matter, effects in materials under irradiation, materials response, related to component lifetime and performance, simulation techniques, and methods of mitigating the impact of radiation on the components and environment in challenging current and future applications.

High and rapid hydrogen release from thermolysis of ammonia borane near PEM fuel cell operating temperature

Science.gov (United States)

Varma, Arvind; Hwang, Hyun Tae; Al-Kukhun, Ahmad

2016-11-15

A system for generating and purifying hydrogen. To generate hydrogen, the system includes inlets configured to receive a hydrogen carrier and an inert insulator, a mixing chamber configured to combine the hydrogen carrier and the inert insulator, a heat exchanger configured to apply heat to the mixture of hydrogen carrier and the inert insulator, wherein the applied heat results in the generation of hydrogen from the hydrogen carrier, and an outlet configured to release the generated hydrogen. To purify hydrogen, the system includes a primary inlet to receive a starting material and an ammonia filtration subassembly, which may include an absorption column configured to absorb the ammonia into water for providing purified hydrogen at a first purity level. The ammonia filtration subassembly may also include an adsorbent member configured to adsorb ammonia from the starting material into an adsorbent for providing purified hydrogen at a second purity level.

Design and manufacturing of the CFRP lightweight telescope structure

Science.gov (United States)

Stoeffler, Guenter; Kaindl, Rainer

2000-06-01

Design of earthbound telescopes is normally based on conventional steel constructions. Several years ago thermostable CFRP Telescope and reflector structures were developed and manufacturing for harsh terrestrial environments. The airborne SOFIA TA requires beyond thermostability an excessive stiffness to mass ratio for the structure fulfilling performance and not to exceed mass limitations by the aircraft Boeing 747 SP. Additional integration into A/C drives design of structure subassemblies. Thickness of CFRP Laminates, either filament wound or prepreg manufactured need special attention and techniques to gain high material quality according to aerospace requirements. Sequential shop assembly of the structure subassemblies minimizes risk for assembling TA. Design goals, optimization of layout and manufacturing techniques and results are presented.

Research program Integrity of Components (FKS). A substantial contribution to component safety

International Nuclear Information System (INIS)

Kussmaul, K.; Roos, E.; Foehl, J.

1998-01-01

The main objectives pursued are: (a) verify the quality of reactor pressure vessels in existing LWR-type reactors, and (b) quantify the safety margin using both specified and non-specified materials and welds. On the basis of knowledge obtained through earlier programmes, the research project was to examine in particular deviations from the specified materials properties, for more exact quantification of the safety margin before RPV failure. There are three major factors influencing the component performance until failure, which are aggregate material fatigue, flaws, loading conditions, and the research work was to focus on the materials properties. An item of main interest was to assess the impact of long service life on the materials properties, assuming particularly unfavourable boundary conditions for materials properties and operational loads. (orig./CB) [de

ITER plasma facing components

International Nuclear Information System (INIS)

Kuroda, T.; Vieider, G.; Akiba, M.

1991-01-01

This document summarizes results of the Conceptual Design Activities (1988-1990) for the International Thermonuclear Experimental Reactor (ITER) project, namely those that pertain to the plasma facing components of the reactor vessel, of which the main components are the first wall and the divertor plates. After an introduction and an executive summary, the principal functions of the plasma-facing components are delineated, i.e., (i) define the low-impurity region within which the plasma is produced, (ii) absorb the electromagnetic radiation and charged-particle flux from the plasma, and (iii) protect the blanket/shield components from the plasma. A list of critical design issues for the divertor plates and the first wall is given, followed by discussions of the divertor plate design (including the issues of material selection, erosion lifetime, design concepts, thermal and mechanical analysis, operating limits and overall lifetime, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, and advanced divertor concepts) and the first wall design (armor material and design, erosion lifetime, overall design concepts, thermal and mechanical analysis, lifetime and operating limits, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, an alternative first wall design, and the limiters used instead of the divertor plates during start-up). Refs, figs and tabs

The determination of contribution of emotional intelligence and parenting styles components to predicts positive psychological components

OpenAIRE

hosein Ebrahimi moghadam; mahin Fekraty

2015-01-01

Background: Since the essential of positive psychological components, as compliment of deficiency oriented approaches, has begun in recent days,we decided to take into account this new branch of psychology which scientifically considers studying forces of human, as well as because of the importance of this branch of psychology, we also tried to search the contribution of emotional intelligence and parenting styles components to predict positive psychological components. Materials and Methods:...

Heavy steel casting components for power plants 'mega-components' made of high Cr-steels

Energy Technology Data Exchange (ETDEWEB)

Hanus, Reinhold [voestalpine Giesserei Linz GmbH, Linz (Austria)

2010-07-01

Steel castings of creep resistant steels play a key role in fossil fuel fired power plants for highly loaded components in the high and intermediate pressure section of the turbines. Inner and outer casings, valve casings, inlet connections and elbows are examples of such critical components. The most important characteristic in a power plant is the efficiency, which mainly drives the CO2-emission. As a consequence of steadily improving power plant efficiencies and ever stricter emission standards, steam parameters become more critical and the creep resistance of the cast materials must also be constantly improved. The foundries voestalpine Giesserei Linz and voestalpine Giesserei Traisen participated in the development of the new 9-10% Cr-steels for application up to 625 C/650 C and in the THERMIE project where Ni-base alloys for 700 C-power plants were developed. Beside the material development in the European research projects the commercial production had to be established for industrial processes and the newly developed materials have to be transferred from research into the commercial production of heavy cast components. After selecting the most promising alloy from the laboratory melts, welding tests were performed - mostly with matching electrodes also produced within COST/THERMIE. Base material and welds were investigated in respect of microstructure, creep resistance, mechanical properties and weldability. Heat treatment investigations were also necessary for optimization of the mechanical properties. Based on the results of these studies, pilot components and plates for testing welding processes were cast in order to verify the castability and weldability of larger parts and to make any necessary adjustments to chemical composition, heat treatment or welding parameters. Parallel to the ongoing creep tests within COST/THERMIE-program, the newly developed steel grades were introduced into the commercial production of large components. This involved finding

Components production and assemble of the irradiation capsule of the Surveillance Program of Materials of the nuclear power plant of Laguna Verde

International Nuclear Information System (INIS)

Medrano, A.

2009-01-01

To predict the effects of the neutrons radiation and the thermal environment about the mechanical properties of the reactor vessel materials of the nuclear power plant of Laguna Verde, a surveillance program is implemented according to the outlines settled by Astm E185-02 -Standard practice for design of surveillance programs for light-water moderated nuclear power reactor vessels-. This program includes the installation of three irradiation capsules of similar materials to those of the reactor vessels, these samples are test tubes for mechanical practices of impact and tension. In the National Institute of Nuclear Research and due to the infrastructure as well as of the actual human resources of the Pilot Plant of Nuclear Fuel Assembles Production it was possible to realize the materials rebuilding extracted in 2005 of Unit 2 of nuclear power plant of Laguna Verde as well as the production, assemble and reassignment of the irradiation capsule made in 2006. At the present time the surveillance materials extracted in 2008 of Unit 1 of the nuclear power plant of Laguna Verde are reconstituting and the components are manufactured for the assembles of the irradiation capsule that will be reinstalled in the reactor vessel in 2010. The purpose of the present work is to describe the necessary components as well as its disposition during the assembles of the irradiation capsule for the surveillance program of the reactors vessel of the nuclear power plant of Laguna Verde. (Author)

Mechanical design of machine components

CERN Document Server

Ugural, Ansel C

2015-01-01

Mechanical Design of Machine Components, Second Edition strikes a balance between theory and application, and prepares students for more advanced study or professional practice. It outlines the basic concepts in the design and analysis of machine elements using traditional methods, based on the principles of mechanics of materials. The text combines the theory needed to gain insight into mechanics with numerical methods in design. It presents real-world engineering applications, and reveals the link between basic mechanics and the specific design of machine components and machines. Divided into three parts, this revised text presents basic background topics, deals with failure prevention in a variety of machine elements and covers applications in design of machine components as well as entire machines. Optional sections treating special and advanced topics are also included.Key Features of the Second Edition:Incorporates material that has been completely updated with new chapters, problems, practical examples...

An assessment of core wide coherency effects in the multichannel modeling of the initiating phase of a severe accident in a sodium fast reactor

International Nuclear Information System (INIS)

Guyot, M.; Gubernatis, P.; Suteau, C.; Le Tellier, R.; Lecerf, J.

2014-01-01

To consolidate the safety assessment for liquid-metal fast breeder reactors (LMFBRs), hypothetical core disruptive accident (HCDA) sequences have been extensively studied over the past decades. Numerous analyses of the so called initiating phase (or primary phase) of a HCDA have been made with the safety analysis system code SAS4A. The SAS4A accident analysis code requires that subassemblies or groups of subassemblies be represented together as independent channels. For simulating a severe accident sequence, a subassembly-to-channel assignment procedure has to be implemented to produce the consistent SAS4A input decks. Generally, one uses imposed criteria over relevant reactor parameters to determine the subassembly to- channel arrangement. The multiple-assembly-per-channel approach introduces core wide coherency effects, which can affect the reactivity balance and therefore the overall accident development. In this paper, a subassembly-to channel assignment procedure based on the subassembly power-to-flow ratio is presented and implemented to generate the SAS4A input decks over a range of parameter values. The corresponding SAS4A calculations have been performed on a large LMFBR. The purpose of the present series of calculations is to investigate the magnitude of errors encountered in the analysis of the initiating phase related to the subassembly-to-channel arrangement selection, by comparison with a one-subassembly-per-channel reference solution. It appears that a refinement in the channel arrangement substantially reduces core wide coherency effects. Analysis of the calculations also suggests that an accurate representation of the scenario requires the number of channels to be on approximately the same order of magnitude as the total number of subassemblies. Numerical results are examined to provide the reader with quantitative measurements of bias related to subassembly to- channel arrangement. (authors)

Printed Spacecraft Separation System

Energy Technology Data Exchange (ETDEWEB)

Dehoff, Ryan R [ORNL; Holmans, Walter [Planetary Systems Corporation

2016-10-01

In this project Planetary Systems Corporation proposed utilizing additive manufacturing (3D printing) to manufacture a titanium spacecraft separation system for commercial and US government customers to realize a 90% reduction in the cost and energy. These savings were demonstrated via “printing-in” many of the parts and sub-assemblies into one part, thus greatly reducing the labor associated with design, procurement, assembly and calibration of mechanisms. Planetary Systems Corporation redesigned several of the components of the separation system based on additive manufacturing principles including geometric flexibility and the ability to fabricate complex designs, ability to combine multiple parts of an assembly into a single component, and the ability to optimize design for specific mechanical property targets. Shock absorption was specifically targeted and requirements were established to attenuate damage to the Lightband system from shock of initiation. Planetary Systems Corporation redesigned components based on these requirements and sent the designs to Oak Ridge National Laboratory to be printed. ORNL printed the parts using the Arcam electron beam melting technology based on the desire for the parts to be fabricated from Ti-6Al-4V based on the weight and mechanical performance of the material. A second set of components was fabricated from stainless steel material on the Renishaw laser powder bed technology due to the improved geometric accuracy, surface finish, and wear resistance of the material. Planetary Systems Corporation evaluated these components and determined that 3D printing is potentially a viable method for achieving significant cost and savings metrics.

Material interactions between system components and glass product melts in a ceramic melter

International Nuclear Information System (INIS)

Knitter, R.

1989-07-01

The interactions of the ceramic and metallic components of a ceramic melter for the vitrification of High Active Waste were investigated with simulated glass product melts in static crucible tests at 1000 0 C and 1150 0 C. Corrosion of the fusion-cast Al 2 O 3 -ZrO 2 -SiO 2 - and Al 2 O 3 -ZrO 2 -SiO 2 -Cr 2 O 3 -refractories (ER 1711 and ER 2161) is characterized by homogeneous chemical dissolution and diffusion through the glass matrix of the refractory. The resulting boundary compositions lead to characteristic modification and formation of phases, not only inside the refractory but also in the glass melt. The attack of the electrode material, a Ni-Cr-Fe-alloy Inconel 690, by the glass melt takes place via grain boundaries and leads to the oxidation of Cr and growth of Cr 2 O 3 -crystals at the boundary layer. Noble metals, added to the glass melt can form solid solutions with the alloy with varying compositions. (orig.) [de

A materials compatibility study in FM-1, a liquid component of a paste extrudable explosive

Energy Technology Data Exchange (ETDEWEB)

Goods, S.H.; Shepodd, T.J.; Mills, B.E. [Sandia National Labs., Livermore, CA (United States); Foster, P. [Mason and Hanger-Silas Mason Co., Inc., Amarillo, TX (United States). Pantex Plant

1993-09-01

The chemical compatibility of various metallic and organic containment materials with a constituent of a paste extrudable explosive (PEX) has been examined through a series of long-term exposures. Corrosion coupons and mechanical test specimens (polymers only) were exposed to FM-1, a principal liquid component of PEX, at 74{degree}C. RX-08-FK is the LLNL designator for this formulation. Compatibility was determined by measuring changes in weight, physical dimensions, and mechanical properties, by examining the coupons for discoloration, surface attack, and corrosion products, and by analyzing for dissolved metals in the FM-1. Of the metals and alloys examined, none of the 300 series stainless steels exhibited adequate corrosion resistance after 74 days of exposure. Copper showed evidence of severe uniform surface attack. Monel 400 also exhibited signs of chemical attack. Nickel and tantalum showed less evidence of attack, although neither, was immune to the liquid. Gold coupons developed a ``tarnish`` film. The gold along with an aluminum alloy, 6061 (in the T6 condition) performed the most satisfactorily. A wide range of polymers were tested for 61 days at 74{degree}C. The materials that exhibited the most favorable response in terms of weight change, dimensional stability, and mechanical properties were Kalrez, PTFE Teflon, and polyethylene.

Insulating jacket for heat sensitive components

International Nuclear Information System (INIS)

Class, G.

1980-01-01

The insulating jacket for long core components of sodium-cooled reactors consists of several layers of austenitic steel, between which a woven wire mesh of the same material is fitted. It is wound in the form of a spiral bandage on the core component. (DG) [de

Materials, critical materials and clean-energy technologies

Science.gov (United States)

Eggert, R.

2017-07-01

Modern engineered materials, components and systems depend on raw materials whose properties provide essential functionality to these technologies. Some of these raw materials are subject to supply-chain risks, and such materials are known as critical materials. This paper reviews corporate, national and world perspectives on material criticality. It then narrows its focus to studies that assess "what is critical" to clean-energy technologies. The focus on supply-chain risks is not meant to be alarmist but rather to encourage attention to monitoring these risks and pursuing technological innovation to mitigate the risks.

Lifetime assessment of service-exposed components

International Nuclear Information System (INIS)

Kalwa, G.; Weber, H.

1988-01-01

A longtime prognosis on the operation of creep-exposed components requires a lifetime analysis. The basis for such an analysis can be improved by an analysis of microstructure and material properties. Actually the grade of material exhaustion has to be regarded as proper assessment quantity. However, stress and time safety also are valuable assessment quantities which should be taken into consideration, especially when the grade of exhaustion is uncertain because of inaccurate input parameters. A correct assessment of the damage state cannot be made without taking into consideration the failure mechanism which has to be assumed for a specific component. With respect to creep the most critical component of a steamline system is the pipe bend because of the risk of large damage events. For this case component metallography by replicas is suggested as preventive test method. The continuation of service of a creep damage pipe bend cannot be recommended. (orig./MM) [de

Evaluation of the physical and electrochemical properties of adobe reinforced and of its component materials

Directory of Open Access Journals (Sweden)

Pérez, G.

2004-06-01

Full Text Available The search of solutions to the habitacional crisis that exists in Latin America has favored to the use of the soil-cement-sisal, adobe reinforced like alternative material of building, in such sense, prevailing to determine the vulnerability of a building of this type, is for that reason that, the characterization of the physical, mechanical and electrochemical properties of the materials that composes it, helps to relate causal the external ones of deterioration to the internal ones. In this particular case it was studied, the permeability, the porosity, the capillaiy absorption, the corrosion potentials (referring to Cu/CuSO₄ and the corrosion rates of adobe reinforced, of its components and their interfaces. In the methodological aspect, permeabilimeter of modified Figg. was used to determine the permeability to the water The porosity and capillary absorption were determined following methods traditional, the corrosion potentials were determined using multimeter and the corrosion rates was made by means of the equipment Gecor 6. Two groups was tried, a first group of the component materials: adobe, mortar for stucco and mortar of reinforcement. The adobe with 5% of cement, mortars for stucco: MFA with 12% of cement and MFB with 16% of cement and internal mortar of reinforcement: MRD, of relation water/cement 0,50. A second group of the composed materials, conformed by 3 specimens test of each one of both types of wall of adobe in where all the individual components are combined: mortar of stucco, adobe and internal mortar: (MFA-Adobe-MRD and (MFB-adobe- MRD. The coefficients of permeability to the water of the component materials oscillate between 5.2.10^-5 for adobe up to 9.71.10^-9 for the internal mortar of reinforcement MRD (a/c=0.50, in mortars for stucco (MFA and MFB this in the 10^-7 order For the case of the composed materials oscillates between 9,38.10^-8 for (MFA-Adobe-MRD until 3

Multicomponent polymeric materials

CERN Document Server

Thomas, Sabu; Saha, Prosenjit

2016-01-01

The book offers an in-depth review of the materials design and manufacturing processes employed in the development of multi-component or multiphase polymer material systems. This field has seen rapid growth in both academic and industrial research, as multiphase materials are increasingly replacing traditional single-component materials in commercial applications. Many obstacles can be overcome by processing and using multiphase materials in automobile, construction, aerospace, food processing, and other chemical industry applications. The comprehensive description of the processing, characterization, and application of multiphase materials presented in this book offers a world of new ideas and potential technological advantages for academics, researchers, students, and industrial manufacturers from diverse fields including rubber engineering, polymer chemistry, materials processing and chemical science. From the commercial point of view it will be of great value to those involved in processing, optimizing an...

Nanostructured composite reinforced material

Science.gov (United States)

Seals, Roland D [Oak Ridge, TN; Ripley, Edward B [Knoxville, TN; Ludtka, Gerard M [Oak Ridge, TN

2012-07-31

A family of materials wherein nanostructures and/or nanotubes are incorporated into a multi-component material arrangement, such as a metallic or ceramic alloy or composite/aggregate, producing a new material or metallic/ceramic alloy. The new material has significantly increased strength, up to several thousands of times normal and perhaps substantially more, as well as significantly decreased weight. The new materials may be manufactured into a component where the nanostructure or nanostructure reinforcement is incorporated into the bulk and/or matrix material, or as a coating where the nanostructure or nanostructure reinforcement is incorporated into the coating or surface of a "normal" substrate material. The nanostructures are incorporated into the material structure either randomly or aligned, within grains, or along or across grain boundaries.

20 years of experience on treatment of large contaminated components and on clearance of material for recycling

International Nuclear Information System (INIS)

Lorenzen, Joachim; Lindberg, Maria; Amcoff, Bjoern; Wirendal, Bo

2005-01-01

This paper will describe the treatment of contaminated, large, retired components from NPP:s, at low and intermediate activity waste levels for recycling in Sweden. Decontamination and melting of various large components, as well as other metal scrap, has been conducted at Studsvik since the mid 1980:ies. Experience on clearance for recycling, i.e. for unconditional re-use of the metals in the public domain will be described. The contaminated material may be Co-60 dominated as well as Uranium Bearing Waste. During these years different techniques for decontamination and segmentation as well as pre- and post treatment have been developed and successively applied at Studsvik melting facility in Nykoeping, Sweden. This collective experience is presently used for the planning and treatment of both domestic and foreign larger components, like heat exchangers, reactors vessel heads, turbine parts, steam generators, fuel bottles and Giant boilers. During 2005 one 300 ton full size, 400 m 3 Westinghouse Steam Generator is under treatment using advanced decontamination, segmentation and melting techniques to be applied in a specifically designed and confined environment. The conduction of demonstration projects as well as commercial projects will be explained and described. The Studsvik melting facility is today treating components and scrap metal comprising stainless and carbon steel as well as aluminium, copper, brass and lead. Studsvik RadWaste has licenses for treating not only components from Swedish nuclear facilities but also for processing components from nuclear industries outside Sweden, including temporary import and export within a limited time window for each international project. Direct clearance or clearance after limited decay storage at Studsvik site is possible. The high Recycling Rate is due to optimized production to leave an extremely low percentage of secondary waste, including post-treatment of the secondary waste volume. Further, the waste volume

Crack growth determination on laboratory components

International Nuclear Information System (INIS)

Hurst, R.C.

1993-01-01

In order to aid design and support remanent life assessment of plant components operating at elevated temperatures, the reliability of the analytical methods, which translate materials data procured from the laboratory to the behaviour of actual components, requires validation. Such a validation can of course be interpreted from operating plant, however the potential risks involved encourage the development of out of plant techniques for the validation of representative components. For meaningful validation, these techniques need careful control and high accuracy which can best be achieved in a laboratory environment. As the laboratory component test should be designed to simulate actual plant conditions as closely as possible, the direct extension of the results to the plant component case requires scaling up. Consequently the successful development of such a test may even lead to the advantageous situation where it could form an alternative to the conventional route where, for example, it may not be possible to obtain the plant component's metallurgical structure in a conventional specimen or, alternatively, when too many assumptions are required in the analysis when translating to different geometries and stress systems. Under these conditions, in spite of the more sophisticated test requirements, it may prove more reasonable to opt for the more representative laboratory component data for use in design or lifetime prediction. The present work describes the application of the component validation test philosophy to the problem of crack growth under two rather different loading conditions. In both cases, crack growth is measured using the direct current potential drop (PD) technique on tubular metallic components containing artificial defects, however the plant conditions to be simulated lead to either creep or thermal fatigue. The creep studies on Alloy 800H support heat exchanger design for nuclear plant, solar towers and chemical plant, whereas the work on the

The monolithic multicell: a tool for testing material components in dye-sensitized solar cells

Energy Technology Data Exchange (ETDEWEB)

Pettersson, H.; Gruszecki, T. [IVF Industrial Research and Development Corporation, Moelndal (Sweden); Bernhard, R. [IVF Industrial Research and Development Corporation, Moelndal (Sweden); The Royal Institute of Technology, Stockholm (Sweden). Center of Molcular Devices, Department of Chemistry; Haeggman, L.; Gorlov, M.; Boschloo, G.; Edvinsson, T.; Kloo, L.; Hagfeldt, A. [The Royal Institute of Technology, Stockholm (Sweden). Center of Molcular Devices, Department of Chemistry

2006-07-01

A multicell is presented as a tool for testing material components in encapsulated dye-sensitized solar cells. The multicell is based on a four-layer monolithic cell structure and an industrial process technology. Each multicell plate includes 24 individual well-encapsulated cells. A sulfur lamp corrected to the solar spectrum has been used to characterize the cells. Efficiencies up to 6.8% at a light-intensity of 1000 W/m{sup su2} (up to 7.5% at 250 W/m{sup 2}) have been obtained with an electrolyte solution based on {upsilon}-butyrolactone. Additionally, a promising long-term stability at cell efficiencies close to 5% at 1000 W/m{sup 2} has been obtained with an electrolyte based on glutaronitrile. The reproducibility of the cell performance before and after exposure to accelerated testing has been high. This means that the multicell can be used as an efficient tool for comparative performance and stability tests. (author)

Critical plasma-wall interaction issues for plasma-facing materials and components in near-term fusion devices

International Nuclear Information System (INIS)

Federici, G.; Coad, J.P.; Haasz, A.A.; Janeschitz, G.; Noda, N.; Philipps, V.; Roth, J.; Skinner, C.H.; Tivey, R.; Wu, C.H.

2000-01-01

The increase in pulse duration and cumulative run-time, together with the increase of the plasma energy content, will represent the largest changes in operation conditions in future fusion devices such as the International Thermonuclear Experimental Reactor (ITER) compared to today's experimental facilities. These will give rise to important plasma-physics effects and plasma-material interactions (PMIs) which are only partially observed and accessible in present-day experiments and will open new design, operation and safety issues. For the first time in fusion research, erosion and its consequences over many pulses (e.g., co-deposition and dust) may determine the operational schedule of a fusion device. This paper identifies the most critical issues arising from PMIs which represent key elements in the selection of materials, the design, and the optimisation of plasma-facing components (PFCs) for the first-wall and divertor. Significant advances in the knowledge base have been made recently, as part of the R and D supporting the engineering design activities (EDA) of ITER, and some of the most relevant data are reviewed here together with areas where further R and D work is urgently needed

A review of phase change materials for vehicle component thermal buffering

International Nuclear Information System (INIS)

Jankowski, Nicholas R.; McCluskey, F. Patrick

2014-01-01

Highlights: • A review of latent heat thermal energy storage for vehicle thermal load leveling. • Examined vehicle applications with transient thermal profiles from 0 to 800 °C. • >700 materials from over a dozen material classes examined for the applications. • Recommendations made for future application of high power density materials. - Abstract: The use of latent heat thermal energy storage for thermally buffering vehicle systems is reviewed. Vehicle systems with transient thermal profiles are classified according to operating temperatures in the range of 0–800 °C. Thermal conditions of those applications are examined relative to their impact on thermal buffer requirements, and prior phase change thermal enhancement studies for these applications are discussed. In addition a comprehensive overview of phase change materials covering the relevant operating range is given, including selection criteria and a detailed list of over 700 candidate materials from a number of material classes. Promising material candidates are identified for each vehicle system based on system temperature, specific and volumetric latent heat, and thermal conductivity. Based on the results of previous thermal load leveling efforts, there is the potential for making significant improvements in both emissions reduction and overall energy efficiency by further exploration of PCM thermal buffering on vehicles. Recommendations are made for further material characterization, with focus on the need for improved data for metallic and solid-state phase change materials for high energy density applications

Surveillance test of the JMTR core components

International Nuclear Information System (INIS)

Takeda, Takashi; Amezawa, Hiroo; Tobita, Kenji

1986-02-01

Surveillance test for the core components of Japan Materials Testing Reactor (JMTR) was started in 1966, and completed in 1985 without one capsule. Most of capsules in the program, except one beryllium specimens, were removed from the core, and carred out the post-irradiation tests at the JMTR Hot Laboratory. The data is applied to review of JMTR core components management plan. JMTR surveillance test was carried out with several kind of materials of JMTR core components, Berylium as the reflector, Hafnium as the neutron absorber of control rod, 17-4PH stainless steel as a roller spring of the control rod, and 304 stainless steel as the grid plate. Results are described in this report. (author)

Integrated Predictive Tools for Customizing Microstructure and Material Properties of Additively Manufactured Aerospace Components

Energy Technology Data Exchange (ETDEWEB)

Radhakrishnan, Balasubramaniam [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fattebert, Jean-Luc [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gorti, Sarma B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Haxhimali, Timor [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); El-Wardany, Tahany [United Technologies Research Center (UTRC), East Hartford, CT (United States); Acharya, Ranadip [United Technologies Research Center (UTRC), East Hartford, CT (United States); Staroselsky, Alexander [United Technologies Research Center (UTRC), East Hartford, CT (United States)

2017-12-01

Additive Manufacturing (AM) refers to a process by which digital three-dimensional (3-D) design data is converted to build up a component by depositing material layer-by-layer. United Technologies Corporation (UTC) is currently involved in fabrication and certification of several AM aerospace structural components made from aerospace materials. This is accomplished by using optimized process parameters determined through numerous design-of-experiments (DOE)-based studies. Certification of these components is broadly recognized as a significant challenge, with long lead times, very expensive new product development cycles and very high energy consumption. Because of these challenges, United Technologies Research Center (UTRC), together with UTC business units have been developing and validating an advanced physics-based process model. The specific goal is to develop a physics-based framework of an AM process and reliably predict fatigue properties of built-up structures as based on detailed solidification microstructures. Microstructures are predicted using process control parameters including energy source power, scan velocity, deposition pattern, and powder properties. The multi-scale multi-physics model requires solution and coupling of governing physics that will allow prediction of the thermal field and enable solution at the microstructural scale. The state-of-the-art approach to solve these problems requires a huge computational framework and this kind of resource is only available within academia and national laboratories. The project utilized the parallel phase-fields codes at Oak Ridge National Laboratory (ORNL) and Lawrence Livermore National Laboratory (LLNL), along with the high-performance computing (HPC) capabilities existing at the two labs to demonstrate the simulation of multiple dendrite growth in threedimensions (3-D). The LLNL code AMPE was used to implement the UTRC phase field model that was previously developed for a model binary alloy, and

Comparative study of material loss at the taper interface in retrieved metal-on-polyethylene and metal-on-metal femoral components from a single manufacturer.

Science.gov (United States)

Bills, Paul; Racasan, Radu; Bhattacharya, Saugatta; Blunt, Liam; Isaac, Graham

2017-08-01

There have been a number of reports on the occurrence of taper corrosion and/or fretting and some have speculated on a link to the occurrence of adverse local tissue reaction specifically in relation to total hip replacement which have a metal-on-metal bearing. As such a study was carried out to compare the magnitude of material loss at the taper in a series of retrieved femoral heads used in metal-on-polyethylene bearings with that in a series of retrieved heads used in metal-on-metal bearings. A total of 36 metal-on-polyethylene and 21 metal-on-metal femoral components were included in the study all of which were received from a customer complaint database. Furthermore, a total of nine as-manufactured femoral components were included to provide a baseline for characterisation. All taper surfaces were assessed using an established corrosion scoring method and measurements were taken of the female taper surface using a contact profilometry. In the case of metal-on-metal components, the bearing wear was also assessed using coordinate metrology to determine whether or not there was a relationship between bearing and taper material loss in these cases. The study found that in this cohort the median value of metal-on-polyethylene taper loss was 1.25 mm 3 with the consequent median value for metal-on-metal taper loss being 1.75 mm 3 . This study also suggests that manufacturing form can result in an apparent loss of material from the taper surface determined to have a median value of 0.59 mm 3 . Therefore, it is clear that form variability is a significant confounding factor in the measurement of material loss from the tapers of femoral heads retrieved following revision surgery.

Materials, critical materials and clean-energy technologies

Directory of Open Access Journals (Sweden)

Eggert R.

2017-01-01

Full Text Available Modern engineered materials, components and systems depend on raw materials whose properties provide essential functionality to these technologies. Some of these raw materials are subject to supply-chain risks, and such materials are known as critical materials. This paper reviews corporate, national and world perspectives on material criticality. It then narrows its focus to studies that assess “what is critical” to clean-energy technologies. The focus on supply-chain risks is not meant to be alarmist but rather to encourage attention to monitoring these risks and pursuing technological innovation to mitigate the risks.

Compatibility study of trans-1,4,5,8-tetranitro-1,4,5,8-tetraazadecalin (TNAD) with some energetic components and inert materials

International Nuclear Information System (INIS)

Yan Qilong; Li Xiaojiang; Zhang Laying; Li Jizhen; Li Hongli; Liu Ziru

2008-01-01

The compatibility of trans-1,4,5,8-tetranitro-1,4,5,8-tetraazadecalin (TNAD) with some energetic components and inert materials of solid propellants was studied by using the pressure DSC method where, cyclotetramethylenetetranitroamine (HMX), cyclotrimethylenetrinitramine (RDX), 1,4-dinitropiperazine (DNP), 1.25/1-NC/NG mixture, lead 3-nitro-1,2,4-triazol-5-onate (NTO-Pb), aluminum powder (Al, particle size = 13.6 μm) and N-nitrodihydroxyethylaminedinitrate (DINA) were used as energetic components and polyethylene glycol (PEG), polyoxytetramethylene-co-oxyethylene (PET), addition product of hexamethylene diisocyanate and water (N-100), 2-nitrodianiline (2-NDPA), 1,3-dimethyl-1,3-diphenyl urea (C 2 ), carbon black (C.B.), aluminum oxide (Al 2 O 3 ), cupric 2,4-dihydroxy-benzoate (β-Cu), cupric adipate (AD-Cu) and lead phthalate (φ-Pb) were used as inert materials. It was concluded that the binary systems of TNAD with NTO-Pb, RDX, PET and Al powder are compatible, and systems of TNAD with DINA and HMX are slightly sensitive, and with 2-NDPA, φ-Pb, β-Cu, AD-Cu and Al 2 O 3 are sensitive, and with PEG, N-100, C 2 and C.B. are incompatible. The impact and friction sensitivity data of the TNAD and TNAD in combination with the other energetic materials under present study was also obtained, and there was no consequential affiliation between sensitivity and compatibility

Development of JOYO MK-II core characteristics database

International Nuclear Information System (INIS)

Tabuchi, Shiro; Aoyama, Takafumi

2000-01-01

The MK-II core of the experimental fast reactor JOYO served as the irradiation bed for testing fuels and materials for FBR development since 1982 for 15 years. During the MK-II operation, extensive data were accumulated from the core management calculations and characteristics tests conducted in thirty-one duty operations and thirteen special test operations. These core management data and core characteristics data were compiled into a database recorded on CD-ROM for user convenience. The calculated core management data are the text style data. The 'Configuration Data' include the history of the fuel exchange and core arrangement for each cycle. The Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of about 300 fuel subassemblies, and 60 irradiation subassemblies. The 'Output Data' include the neutron fluxes, gamma fluxes, power density, linear heat rates, coolant and fuel temperature distributions of each core position at the beginning and end of each cycle. The measured core characteristics data, such as the excess reactivity, control rod worths, temperature coefficient, power coefficient, and burn-up coefficient are also included along with the measurement conditions. (J.P.N.)

Metallurgical Laboratory and Components Testing

Data.gov (United States)

Federal Laboratory Consortium — In the field of metallurgy, TTC is equipped to run laboratory tests on track and rolling stock components and materials. The testing lab contains scanning-electron,...

Special purpose materials for fusion application

International Nuclear Information System (INIS)

Scott, J.L.; Clinard, F.W. Jr.; Wiffen, F.W.

1984-01-01

Originally in 1978 the Special Purpose Materials Task Group was concerned with tritium breeding materials, coolants, tritium barriers, graphite and silicon carbide, ceramics, heat-sink materials, and magnet components. Since then several other task groups have been created, so now the category includes only materials for superconducting magnets and ceramics. For the former application copper-stabilized Nb 3 Sn (Ti) insulated with polyimides will meet the general requirements, so that testing of prototype components is the priority task. Ceramics are required for several critical components of fusion reactors either as dielectrics or as a structural material. Components near the first wall will receive exposures of 5 to 20 MW.year/m"2. Other ceramic applications are well behind the first wall, with lower damage levels. Most insulators operate near room temperature, but ceramic blanket structures may operate up to 1000 0 C. Because of a meager data base, one cannot identify optimum ceramics for structural application; but MgAl 2 O 4 is an attractive dielectric material

The role of damage analysis in the assessment of service-exposed components

International Nuclear Information System (INIS)

Bendick, W.; Muesch, H.; Weber, H.

1987-01-01

Components in power stations are subjected to service conditions under which creep processes take place limiting the component's lifetime by material exhaustion. To ensure a safe and economic plant operation it is necessary to get information about the exhaustion grade of single components as well as of the whole plant. A comprehensive lifetime assessment requests the complete knowledge of the service parameters, the component's deformtion behavior, and the change in material properties caused by longtime exposure to high service temperatures. A basis of evaluation is given by: 1) determination of material exhaustion by calculation, 2) investigation of the material properties, and 3) damage analysis. The purpose of this report is to show the role which damage analysis can play in the assessment of service-exposed components. As an example the test results of a damaged pipe bend will be discussed. (orig./MM)

Modulating the level of components within plants

Science.gov (United States)

Bobzin, Steven Craig; Apuya, Nestor; Chiang, Karen; Doukhanina, Elena; Feldmann, Kenneth; Jankowski, Boris; Margolles-Clark, Emilio; Mumenthaler, Daniel; Okamuro, Jack; Park, Joon-Hyun; Van Fleet, Jennifer E.; Zhang, Ke

2017-09-12

Materials and Methods for identifying lignin regulatory region-regulatory protein associations are disclosed. Materials and methods for modulating lignin accumulation are also disclosed. In addition, methods and materials for modulating (e.g., increasing or decreasing) the level of a component (e.g., protein, oil, lignin, carbon, a carotenoid, or a triterpenoid) in plants are disclosed.

Electronic components and systems

CERN Document Server

Dennis, W H

2013-01-01

Electronic Components and Systems focuses on the principles and processes in the field of electronics and the integrated circuit. Covered in the book are basic aspects and physical fundamentals; different types of materials involved in the field; and passive and active electronic components such as capacitors, inductors, diodes, and transistors. Also covered in the book are topics such as the fabrication of semiconductors and integrated circuits; analog circuitry; digital logic technology; and microprocessors. The monograph is recommended for beginning electrical engineers who would like to kn

Thermal fatigue equipment to test joints of materials for high heat flux components

International Nuclear Information System (INIS)

Visca, E.; Libera, S.; Orsini, A.; Riccardi, B.; Sacchetti, M.

2000-01-01

The activity, carried out in the framework of an ITER divertor task, was aimed at defining a suitable method in order to qualify junctions between armour materials and heat sink of plasma-facing components (PFCs) mock-ups. An equipment able to perform thermal fatigue testing by electrical heating and active water-cooling was constructed and a standard for the sample was defined. In this equipment, during operation cycles, two samples are heated by thermal contact up to a relevant temperature value (350 deg. C) and then the water flow is switched on, thus producing fast cooling with time constants and gradients close to the real operating conditions. The equipment works with a test cycle of about 60 s and is suitable for continuous operation. A complete test consists of about 10000 cycles. After the assembling, the equipment and the control software were optimized to obtain a good reliability. Preliminary tests on mock-ups with flat CFC tiles joined to copper heat sink were performed. Finite-elements calculations were carried out in order to estimate the value of the thermal stresses arising close to the joint under the transient conditions that are characteristic of this equipment

Advanced tungsten materials for plasma-facing components of DEMO and fusion power plants

International Nuclear Information System (INIS)

Neu, R.; Riesch, J.; Coenen, J.W.; Brinkmann, J.; Calvo, A.; Elgeti, S.; García-Rosales, C.; Greuner, H.; Hoeschen, T.; Holzner, G.; Klein, F.; Koch, F.

2016-01-01

Highlights: • Development of W-fibre enhanced W-composites incorporating extrinsic toughening mechanisms. • Production of a large sample (more than 2000 long fibres) for mechanical and thermal testing. • Even in a fully embrittled state, toughening mechanisms are still effective. • Emissions of volatile W-oxides can be suppressed by alloying W with elements forming stable oxides. • WCr10Ti2 has been successfully tested under accidental conditions and high heat fluxes. - Abstract: Tungsten is the major candidate material for the armour of plasma facing components in future fusion devices. To overcome the intrinsic brittleness of tungsten, which strongly limits its operational window, a W-fibre enhanced W-composite material (W_f/W) has been developed incorporating extrinsic toughening mechanisms. Small W_f/W samples show a large increase in toughness. Recently, a large sample (50 mm × 50 mm × 3 mm) with more than 2000 long fibres has been successfully produced allowing further mechanical and thermal testing. It could be shown that even in a fully embrittled state, toughening mechanisms as crack bridging by intact fibres, as well as the energy dissipation by fibre-matrix interface debonding and crack deflection are still effective. A potential problem with the use of pure W in a fusion reactor is the formation of radioactive and highly volatile WO_3 compounds and their potential release under accidental conditions. It has been shown that the oxidation of W can be strongly suppressed by alloying with elements forming stable oxides. WCr10Ti2 alloy has been produced on a technical scale and has been successfully tested in the high heat flux test facility GLADIS. Recently, W-Cr-Y alloys have been produced on a lab-scale. They seem to have even improved properties compared to the previously investigated W alloys.

Advanced tungsten materials for plasma-facing components of DEMO and fusion power plants

Energy Technology Data Exchange (ETDEWEB)

Neu, R., E-mail: Rudolf.Neu@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Fakultät für Maschinenbau, Technische Universität München, D-85748 Garching (Germany); Riesch, J. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Coenen, J.W. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Brinkmann, J. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Calvo, A. [CEIT and Tecnun (University of Navarra), E-20018 San Sebastian (Spain); Elgeti, S. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); García-Rosales, C. [CEIT and Tecnun (University of Navarra), E-20018 San Sebastian (Spain); Greuner, H.; Hoeschen, T.; Holzner, G. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Klein, F. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Koch, F. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); and others

2016-11-01

Highlights: • Development of W-fibre enhanced W-composites incorporating extrinsic toughening mechanisms. • Production of a large sample (more than 2000 long fibres) for mechanical and thermal testing. • Even in a fully embrittled state, toughening mechanisms are still effective. • Emissions of volatile W-oxides can be suppressed by alloying W with elements forming stable oxides. • WCr10Ti2 has been successfully tested under accidental conditions and high heat fluxes. - Abstract: Tungsten is the major candidate material for the armour of plasma facing components in future fusion devices. To overcome the intrinsic brittleness of tungsten, which strongly limits its operational window, a W-fibre enhanced W-composite material (W{sub f}/W) has been developed incorporating extrinsic toughening mechanisms. Small W{sub f}/W samples show a large increase in toughness. Recently, a large sample (50 mm × 50 mm × 3 mm) with more than 2000 long fibres has been successfully produced allowing further mechanical and thermal testing. It could be shown that even in a fully embrittled state, toughening mechanisms as crack bridging by intact fibres, as well as the energy dissipation by fibre-matrix interface debonding and crack deflection are still effective. A potential problem with the use of pure W in a fusion reactor is the formation of radioactive and highly volatile WO{sub 3} compounds and their potential release under accidental conditions. It has been shown that the oxidation of W can be strongly suppressed by alloying with elements forming stable oxides. WCr10Ti2 alloy has been produced on a technical scale and has been successfully tested in the high heat flux test facility GLADIS. Recently, W-Cr-Y alloys have been produced on a lab-scale. They seem to have even improved properties compared to the previously investigated W alloys.

Cladding Effects on Structural Integrity of Nuclear Components

International Nuclear Information System (INIS)

Sattari-Far, Iradi; Andersson, Magnus

2006-06-01

Based on this study, the following conclusions and recommendations can be made: Due to significant differences in the thermal and mechanical properties between the austenitic cladding and the ferritic base metal, residual stresses are induced in the cladding and the underlying base metal. These stresses are left in clad components even after Post-Weld Heat Treatment (PWHT). The different restraint conditions of the clad component have a minor influence on the magnitude of the cladding residual stresses in the cladding layer. The thickness of the clad object is the main impacting geometrical dimension in developing cladding residual stresses. A clad object having a base material thickness exceeding 10 times the cladding thickness would be practically sufficient to introduce cladding residual stresses of a thick reactor pressure vessel. For a clad component that received PWHT, the peak tensile stress is in the cladding layer, and the residual stresses in the underlying base material are negligible. However, for clad components not receiving PWHT, for instance the repair welding of the cladding, the cladding residual stresses of tensile type exist even in the base material. This implies a higher risk for underclad cracking for clad repairs that received no PWHT. For certain clad geometries, like nozzles, the profile of the cladding residual stresses depends on the clad thickness and position, and significant tensile stresses can also exist in the base material. Based on different measurements reported in the literature, a value of 150 GPa can be used as Young's Modulus of the austenitic cladding material at room temperature. The control measurements of small samples from the irradiated reactor pressure vessel head did not reveal a significant difference of Young's Modulus between the irradiated and the unirradiated cladding material condition. No significant differences between the axial and tangential cladding residual stresses are reported in the measurement of

Flight service evaluation of composite helicopter components

Science.gov (United States)

Mardoian, George H.; Ezzo, Maureen B.

1994-01-01

This paper presents the results of a NASA funded contract and Sikorsky research and development programs to evaluate structural composite components in flight service on Sikorsky Model S-76 helicopters. Selected components were removed and tested at prescribed intervals over a nine year time frame. Four horizontal stabilizers and thirteen tail rotor spars were returned from commercial service in West Palm Beach, Florida and in the Gulf Coast region of Louisiana to determine the long term effects of operations in hot and humid climates on component performance. Concurrent with the flight component evaluation, panels of materials used in their fabrication were exposed to the environment in ground racks. Selected panels were tested annually to determine the effects of exposure on physical and mechanical properties. The results of 55,741 component flight hours and 911 months of field exposure are reported and compared with initial Federal Aviation Administration (FAA) certification data. The findings of this program have provided increased confidence in the long term durability of advanced composite materials used in helicopter structural applications.

Mechanical and thermal resistance of multi-material components for ITER

International Nuclear Information System (INIS)

Burlet, H.

2013-01-01

The First Wall panels for ITER are complex parts composed of stainless steel, copper and beryllium [1]. These materials are joined using diffusion bonding technique. The stainless steel is a commonly used in nuclear reactors 316LN material and acts as a structural material. The copper alloy is a CuCrZr material which acts as a heat sink. The beryllium consisting in tiles and layer is used as the protective plasma facing material. The fabrication of these panels is performed through 2 main steps. The first step consists in welding all together a bi-metallic support structure made from a thick CuCrZr plate embedded with 316LN tubes and bonded to a thick 316LN backing plate with cooling channels. The bonding is performed in a HIP (Hot Isostatic Pressure) facility. The second step is performed at a lower temperature and aims at simultaneously welding by HIP Be onto CuCrZr and ageing the CuCrZr heat sink to obtain the correct mechanical resistance of this alloy reinforced by precipitates. The various joints 316LN/316LN, 316LN/CuCrZr, and CuCrZr/Be are then characterized [2] from a microstructural point of view and by mechanical tests. It is quite hard to characterize the strength of a diffusion bonded joints. Standard tests may be used for homogeneous joints whereas specific tests have been developed to characterize the heterogeneous bonds. To optimize the bond, we performed mainly impact and tensile bi-material tests (Fig 1). Once the manufacture parameters have been optimized, advanced mechanical tests are performed based on Bimetallic CT specimens, axisymmetric notched specimens, 4P bending specimens. Numerical simulations are required to analyse the mechanical response. In order to characterize the fatigue resistance of the joints, specific mock-ups have been designed by the European Fusion Development Agreement EFDA team (Fig 2). Results of heat flux testing will be reviewed for the various joints. The assembly of heterogeneous materials by Hipping is very complex

MENT reconstruction and potting comparison of a LMFBR fuel bundle

International Nuclear Information System (INIS)

Betten, P.R.; Tow, D.M.

1984-01-01

Since the advent of computer-assisted-tomography (CAT), the CAT techniques have been rapidly expanded to the nuclear industry. A number of investigators have applied these techniques to reconstruct the fuel bundle configuration inside a subassembly with various degrees of resolution; however, there has been little data available on the accuracy of these reconstructions, and no comparisons have been made with the internal structure of actual irradiated subassemblies. Some efforts have utilized pretest mock-ups to calibrate the CAT algorithms, but the resulting mock-up configurations do not necessarily represent an actual subassembly, so an exact comparison has been lacking. The purpose of this paper is to present the results of a comparison between a CAT reconstruction of an irradiated subassembly and the destructive examination of the same subassembly

On a Thermodynamic Approach to Material Selection for Service in Aggressive Multi-Component Gaseous and/or Vapor Environments

Energy Technology Data Exchange (ETDEWEB)

Glazoff, Michael Vasily [Idaho National Lab. (INL), Idaho Falls, ID (United States); Marschman, Steven Craig [Idaho National Lab. (INL), Idaho Falls, ID (United States); Soelberg, Nicholas Ray [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-09-01

This report fulfills the M4 milestone, M4FT-15IN08020110 UNF Analysis Support, under Work Package Number FT-15IN080201. The issue of materials selection for many engineering applications represents an important problem, particularly in cases where material failure is possible as a result of corrosive environments. For example, 304 dual purpose or 316 stainless steel is used in the construction of many used nuclear fuel storage canisters. Deployed all over the world, these canisters are housed inside shielded enclosures and cooled passively by convective airflow. When located along seaboards or particular industrial areas, salt, other corrosive chemicals, and moisture can become entrained in the air that cools the canisters. It is important to develop an understanding of what impact, if any, that chemical environment will have on those canisters. In many cases of corrosion in aggressive gaseous environments, the material selection process is based on some general recommendations, anecdotal evidence, and/or the past experience of that particular project’s participants. For gaseous mixtures, the theoretical basis is practically limited to the construction of the so-called “Ellingham diagrams” for pure metals. These plots predict the equilibrium temperature between different individual metals, their respective oxides, and oxygen gas. Similar diagrams can be constructed for the reactions with sulfur, nitrogen, carbon, etc. In the generalization of this approach by Richardson and Jeffes, additional scales can be superimposed upon an Ellingham diagram that would correspond to different gaseous mixtures, e.g. CO/CO₂, or H₂/H₂O. However, while the general approach to predicting the stability of a multi-component heterogeneous alloy (e.g., steel or a superalloy) in a multi-component aggressive gaseous environment was developed in very general form, actual examples of its applications to concrete real-life problems are practically absent

Alloy 800 specifications in compliance with component requirements

International Nuclear Information System (INIS)

Diehl, H.; Bodmann, E.

1990-01-01

In view of the importance of the material Alloy 800 in high-temperature reactor plants (HTR), a material data bank was established which is used for statistical evaluation of mechanical and physical material behaviour. Based on investigations on the interconnection between the mechanical properties at high temperatures and the metallurgical parameters, different types of Alloy 800 were specified in compliance with the component requirements. In addition, aspects of corrosion and toughness behaviour were taken into consideration. The specifications and strength characteristics for the different variants of Alloy 800 were incorporated into draft DIN standards after discussion and approval in expert committees. Further important characteristics of the mechanical and physical material behaviour were summarized in HTR material data sheets so as to furnish an improved basis for the design and stress analyses of Alloy 800 components. (orig.)

Analysis of excess reactivity of JOYO MK-III performance test core

International Nuclear Information System (INIS)

Maeda, Shigetaka; Yokoyama, Kenji

2003-10-01

JOYO is currently being upgraded to the high performance irradiation bed JOYO MK-III core'. The MK-III core is divided into two fuel regions with different plutonium contents. To obtain a higher neutron flux, the active core height was reduced from 55 cm to 50 cm. The reflector subassemblies were replaced by shielding subassemblies in the outer two rows. Twenty of the MK-III outer core fuel subassemblies in the performance test core were partially burned in the transition core. Four irradiation test rigs, which do not contain any fuel material, were loaded in the center of the performance test core. In order to evaluate the excess reactivity of MK-III performance test core accurately, we evaluated it by applying not only the JOYO MK-II core management code system MAGI, but also the MK-III core management code system HESTIA, the JUPITER standard analysis method and the Monte Carlo method with JFS-3-J3.2R content set. The excess reactivity evaluations obtained by the JUPITER standard analysis method were corrected to results based on transport theory with zero mesh-size in space and angle. A bias factor based on the MK-II 35th core, which sensitivity was similar to MK-III performance test core's, was also applied, except in the case where an adjusted nuclear cross-section library was used. Exact three-dimensional, pin-by-pin geometry and continuous-energy cross sections were used in the Monte Carlo calculation. The estimated error components associated with cross-sections, methods correction factors and the bias factor were combined based on Takeda's theory. Those independently calculated values agree well and range from 2.8 to 3.4%Δk/kk'. The calculation result of the MK-III core management code system HESTLA was 3.13% Δk/kk'. The estimated errors for bias method range from 0.1 to 0.2%Δk/kk'. The error in the case using adjusted cross-section was 0.3%Δk/kk'. (author)

Beryllium assessment and recommendation for application in ITER plasma facing components

Energy Technology Data Exchange (ETDEWEB)

Barabash, V.; Tanaka, S.; Matera, R. [ITER Joint Central Team, Muenchen (Germany)

1998-01-01

The design status of the ITER Plasma Facing Components (PFC) is presented. The operational conditions of the armour material for the different components are summarized. Beryllium is the reference armour material for the Primary Wall, Baffle and Limiter and the back-up material for the Divertor Dome. The activities on the selection of the Be grades and the joining technologies are reviewed. (author)

Procedures and instrumentation for sodium boiling experiments in EBR-II

International Nuclear Information System (INIS)

Crowe, R.D.

1976-01-01

The development of instrumentation capable of detecting localized coolant boiling in a liquid metal cooled breeder reactor (LMFBR) has a high priority in fast reactor safety. The detection must be rapid enough to allow corrective action to be taken before significant damage occurs to the core. To develop and test a method of boiling detection, it is desirable to produce boiling in a reactor and thereby introduce a condition in the reactor the original design concepts were chosen to preclude. The proposed boiling experiments are designed to safely produce boiling in the subassembly of a fast reactor and provide the information to develop boiling detection instrumentation without core damage or safety compromise. The experiment consists of the operation of two separate subassemblies, first, a gamma heated boiling subassembly which produces non-typical but highly conservative boiling and then a fission heated subassembly which simulates a prototypical boiling event. The two boiling subassemblies are designed to operate in the instrumentation subassembly test facility (INSAT) of Experiment Breeder Reactor II

Methodology for predicting the life of waste-package materials, and components using multifactor accelerated life tests

International Nuclear Information System (INIS)

Thomas, R.E.; Cote, R.W.

1983-09-01

Accelerated life tests are essential for estimating the service life of waste-package materials and components. A recommended methodology for generating accelerated life tests is described in this report. The objective of the methodology is to define an accelerated life test program that is scientifically and statistically defensible. The methodology is carried out using a select team of scientists and usually requires 4 to 12 man-months of effort. Specific agendas for the successive meetings of the team are included in the report for use by the team manager. The agendas include assignments for the team scientists and a different set of assignments for the team statistician. The report also includes descriptions of factorial tables, hierarchical trees, and associated mathematical models that are proposed as technical tools to guide the efforts of the design team

SSME Alternate Turbopump Development Program: Design verification specification for high-pressure fuel turbopump

Science.gov (United States)

1989-01-01

The design and verification requirements are defined which are appropriate to hardware at the detail, subassembly, component, and engine levels and to correlate these requirements to the development demonstrations which provides verification that design objectives are achieved. The high pressure fuel turbopump requirements verification matrix provides correlation between design requirements and the tests required to verify that the requirement have been met.

[Simultaneous separation and detection of principal component isomer and related substances of raw material drug of ammonium glycyrrhizinate by RP-HPLC and structure confirmation].

Science.gov (United States)

Zhao, Yan-Yan; Liu, Li-Yan; Han, Yuan-Yuan; Li, Yue-Qiu; Wang, Yan; Shi, Min-Jian

2013-08-01

A simple, fast and sensitive analytical method for the simultaneous separation and detection of 18alpha-glycyrrhizinic acid, 18beta-glycyrrhizinic acid, related substance A and related substance B by RP-HPLC and drug quality standard was established. The structures of principal component isomer and related substances of raw material drug of ammonium glycyrrhizinate have been confirmed. Reference European Pharmacopoeia EP7.0 version, British Pharmacopoeia 2012 version, National Drug Standards of China (WS 1-XG-2002), domestic and international interrelated literature were referred to select the composition of mobile phase. The experimental parameters including salt concentration, pH, addition quantities of organic solvent, column temperature and flow rate were optimized. Finally, the assay was conducted on a Durashell-C18 column (250 mm x 4.6 mm, 5 microm) with 0.01 mol x mL(-1) ammonium perchlorate (add ammonia to adjust the pH value to 8.2) -methanol (48 : 52) as mobile phase at the flow rate of 0.8 mL x min(-1), and the detection wavelength was set at 254 nm. The column temperature was 50 degrees C and the injection volume was 10 microL. The MS, NMR, UV and RP-HPLC were used to confirm the structures of principal component isomer and related substances of raw material drug of ammonium glycyrrhizinate. Under the optimized separation conditions, the calibration curves of 18 alpha-glycyrrhizinic acid, 18beta-glycyrrhizinic acid, related substance A and related substance B showed good linearity within the concentration of 0.50-100 microg x mL(-1) (r = 0.999 9). The detection limits for 18alpha-glycyrrhizinic acid, 18beta-glycyrrhizinic acid, related substance A and related substance B were 0.15, 0.10, 0.10, 0.15 microg x mL(-1) respectively. The method is sensitive, reproducible and the results are accurate and reliable. It can be used for chiral resolution of 18alpha-glycyrrhizinic acid, 18Pbeta-glycyrrhizinic acid, and detection content of principal component and

Passive Microwave Components and Antennas

DEFF Research Database (Denmark)

State-of-the-art microwave systems always require higher performance and lower cost microwave components. Constantly growing demands and performance requirements of industrial and scientific applications often make employing traditionally designed components impractical. For that reason, the design...... and development process remains a great challenge today. This problem motivated intensive research efforts in microwave design and technology, which is responsible for a great number of recently appeared alternative approaches to analysis and design of microwave components and antennas. This book highlights...... techniques. Modelling and computations in electromagnetics is a quite fast-growing research area. The recent interest in this field is caused by the increased demand for designing complex microwave components, modeling electromagnetic materials, and rapid increase in computational power for calculation...

French R&D on Materials for the Core Components of SFRs

International Nuclear Information System (INIS)

Le Flem, M.; Séran, J.L.; Blat-Yrieix, M.; Garat, V.

2013-01-01

ASTRID demonstrator 480-700°C, 110 dpa. • Use of reference materials benefiniting from a large feed-back from the previous French SFRs (Rapsodie, Phénix, SuperPhénix) • Austenitic steels (cladding), Martensitic steels (wrapper tube), B4C (absorbers). • Improving the description of their behavior (swelling, high temperature) • Qualifying the materials regarding the specificities of ASTRID core. Future SFRs 530-750, 180 dpa. • Use of advanced materials with improved properties • ODS ferritic/martensitic steels (cladding), Other metallic solutions as V alloys (cladding), SiC/SiC composites (wrapper tube), Innovative absorbers and reflectors. • R&D to develop/fabricate suitable grades • Qualifying these materials in ASTRID

Power electronics handbook components, circuits and applications

CERN Document Server

Mazda, F F

1993-01-01

Power Electronics Handbook: Components, Circuits, and Applications is a collection of materials about power components, circuit design, and applications. Presented in a practical form, theoretical information is given as formulae. The book is divided into three parts. Part 1 deals with the usual components found in power electronics such as semiconductor devices and power semiconductor control components, their electronic compatibility, and protection. Part 2 tackles parts and principles related to circuits such as switches; link frequency chargers; converters; and AC line control, and Part 3

High-temperature materials and structural ceramics

International Nuclear Information System (INIS)

1990-01-01

This report gives a survey of research work in the area of high-temperature materials and structural ceramics of the KFA (Juelich Nuclear Research Center). The following topics are treated: (1) For energy facilities: ODS materials for gas turbine blades and heat exchangers; assessment of the remaining life of main steam pipes, material characterization and material stress limits for First-Wall components; metallic and graphitic materials for high-temperature reactors. (2) For process engineering plants: composites for reformer tubes and cracking tubes; ceramic/ceramic joints and metal/ceramic and metal/metal joints; Composites and alloys for rolling bearing and sliding systems up to application temperatures of 1000deg C; high-temperature corrosion of metal and ceramic material; porous ceramic high-temperature filters and moulding coat-mix techniques; electrically conducting ceramic material (superconductors, fuel cells, solid electrolytes); high-temperature light sources (high-temperature chemistry); oil vapor engines with caramic components; ODS materials for components in diesel engines and vehicle gas turbines. (MM) [de

Neutrons and synchrotron radiation in engineering materials science from fundamentals to material and component characterization

CERN Document Server

Reimers, W; Schreyer, A; Clemens, H; Kaysser-Pyzalla, Anke Rita

2008-01-01

Besides its coverage of the four important aspects of synchrotron sources, materials and material processes, measuring techniques, and applications, this ready reference presents both important method types: diffraction and tomography. Following an introduction, a general section leads on to methods, while further sections are devoted to emerging methods and industrial applications. In this way, the text provides new users of large-scale facilities with easy access to an understanding of both the methods and opportunities offered by different sources and instruments.

Euro hybrid materials and structures. Proceedings

Energy Technology Data Exchange (ETDEWEB)

Hausmann, Joachim M.; Siebert, Marc (eds.)

2016-08-01

In order to use the materials as best as possible, several different materials are usually mixed in one component, especially in the field of lightweight design. If these combinations of materials are joined inherently, they are called multi material design products or hybrid structures. These place special requirements on joining technology, design methods and manufacturing and are challenging in other aspects, too. The eight chapters with manuscripts of the presentations are: Chapter 1- Interface: What happens in the interface between the two materials? Chapter 2 - Corrosion and Residual Stresses: How about galvanic corrosion and thermal residual stresses in the contact zone of different materials? Chapter 3 - Characterization: How to characterize and test hybrid materials? Chapter 4 - Design: What is a suitable design and dimensioning method for hybrid structures? Chapter 5 - Machining and Processing: How to machine and process hybrid structures and materials? Chapter 6 - Component Manufacturing: What is a suitable manufacturing route for hybrid structures? Chapter 7 - Non-Destructive Testing and Quality Assurance: How to assure the quality of material and structures? Chapter 8 - Joining: How to join components of different materials?.

Euro hybrid materials and structures. Proceedings

International Nuclear Information System (INIS)

Hausmann, Joachim M.; Siebert, Marc

2016-01-01

In order to use the materials as best as possible, several different materials are usually mixed in one component, especially in the field of lightweight design. If these combinations of materials are joined inherently, they are called multi material design products or hybrid structures. These place special requirements on joining technology, design methods and manufacturing and are challenging in other aspects, too. The eight chapters with manuscripts of the presentations are: Chapter 1- Interface: What happens in the interface between the two materials? Chapter 2 - Corrosion and Residual Stresses: How about galvanic corrosion and thermal residual stresses in the contact zone of different materials? Chapter 3 - Characterization: How to characterize and test hybrid materials? Chapter 4 - Design: What is a suitable design and dimensioning method for hybrid structures? Chapter 5 - Machining and Processing: How to machine and process hybrid structures and materials? Chapter 6 - Component Manufacturing: What is a suitable manufacturing route for hybrid structures? Chapter 7 - Non-Destructive Testing and Quality Assurance: How to assure the quality of material and structures? Chapter 8 - Joining: How to join components of different materials?

EU Development of High Heat Flux Components

International Nuclear Information System (INIS)

Linke, J.; Lorenzetto, P.; Majerus, P.; Merola, M.; Pitzer, D.; Roedig, M.