A study on improvement of analytical prediction model for spacer grid pressure loss coefficients

International Nuclear Information System (INIS)

Lim, Jonh Seon

2002-02-01

Nuclear fuel assemblies used in the nuclear power plants consist of the nuclear fuel rods, the control rod guide tubes, an instrument guide tube, spacer grids,a bottom nozzle, a top nozzle. The spacer grid is the most important component of the fuel assembly components for thermal hydraulic and mechanical design and analyses. The spacer grids fixed with the guide tubes support the fuel rods and have the very important role to activate thermal energy transfer by the coolant mixing caused to the turbulent flow and crossflow in the subchannels. In this paper, the analytical spacer grid pressure loss prediction model has been studied and improved by considering the test section wall to spacer grid gap pressure loss independently and applying the appropriate friction drag coefficient to predict pressure loss more accurately at the low Reynolds number region. The improved analytical model has been verified based on the hydraulic pressure drop test results for the spacer grids of three types with 5x5, 16x16, 17x17 arrays, respectively. The pressure loss coefficients predicted by the improved analytical model are coincident with those test results within ±12%. This result shows that the improved analytical model can be used for research and design change of the nuclear fuel assembly

Impact analysis of the spacer grid assembly and shape optimization of the attached spring

Energy Technology Data Exchange (ETDEWEB)

Park, K. J.; Lee, Z. N. [Hanyang University, Seoul (Korea)

2002-04-01

Spacer grids support fuel rods and maintain geometry from external impact loads. A simulation is performed for the strength of a spacer grid under the impact load. The critical impact load that leads to plastic deformation is identified by a free-fall test. A finite element model is established for the nonlinear simulation of the impact process. The simulation model is tuned based on the free-fall test. The model considers the aspects of welding and the contacts between components. Nonlinear finite element analysis is carried out using a software system called ABAQUS/EXPLICIT. The results are discussed from a design viewpoint. Design requirements are defined and a design process is established. The design process includes mathematical optimization as well as practical design method. The shape of the grid spring is designed to maintain its function during the lifetime of the fuel assembly. A structural optimization method is employed for the shape design. A good design is found. Commercial codes are utilized for structural analysis and optimization. 18 refs., 61 figs., 3 tabs. (Author)

Deformation behavior of cell spring of an irradiated spacer grid

International Nuclear Information System (INIS)

Jin, Y. G.; Baek, S. J.; Ryu, W. S.; Kim, G. S.; Yoo, B. O.; Kim, D. S.; Ahn, S. B.; Chun, Y. B.; Choo, Y. S.

2012-01-01

Mechanical properties of a space grid of a fuel assembly are of great importance for fuel operation reliability in extended fuel burnup and duration of fuel life. The spacer grid with inner and outer straps has cell spring and dimples, which are in contact with the fuel rod. The spacer grids supporting the fuel rods absorb vibration impacts due to the reactor coolant flow and also grid spring force is decreasing under irradiation. This reduction of contact force might cause the grid to rod fretting wear. The fretting failure of the fuel rod is one of the significant issues recently in the nuclear industry from an economical as well as a safety concern. Thus, it is important to understand the characteristics of cell spring behavior for an irradiated spacer grid. In the present study, the stiffness test and dimensional measurement of cell springs were conducted to investigate the deformation behavior of cell springs of an irradiated spacer grid in a hot cell at IMEF (irradiated materials examination facility) of KAERI

5 X 5 rod bundle flow field measurements downstream a PWR spacer grid

Energy Technology Data Exchange (ETDEWEB)

Castro, Higor F.P.; Silva, Vitor V A.; Santos, André A.C.; Veloso, Maria A.F., E-mail: higorfabiano@gmail.com, E-mail: mdora@nuclear.ufmg.br, E-mail: vitors@cdtn.br, E-mail: aacs@cdtn.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil); Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

The spacer grids are structures present in nuclear fuel assembly of Pressurized Water Reactors (PWR). They play an important structural role and also assist in heat removal through the assembly by promoting increased turbulence of the flow. Understanding the flow dynamics downstream the spacer grid is paramount for fuel element design and analysis. This paper presents water flow velocity profiles measurements downstream a spacer grid in a 5 x 5 rod bundle test rig with the objective of highlighting important fluid dynamic behavior near the grid and supplying data for CFD simulation validation. These velocity profiles were obtained at two different heights downstream the spacer grid using a LDV (Laser Doppler Velocimetry) through the top of test rig. The turbulence intensities and patterns of the swirl and cross flow were evaluated. The tests were conducted for Reynolds numbers ranging from 1.8 x 10{sup 4} to 5.4 x 10{sup 4}. This experimental research was carried out in thermo-hydraulics laboratory of Nuclear Technology Development Center – CDTN. Results show great repeatability and low uncertainties (< 1.24 %). Details of the flow field show how the mixture and turbulence induced by the spacer grid quickly decays downstream the spacer grid. It is shown that the developed methodology can supply high resolution low uncertainty results that can be used for validation of CFD simulations. (author)

5 X 5 rod bundle flow field measurements downstream a PWR spacer grid

International Nuclear Information System (INIS)

Castro, Higor F.P.; Silva, Vitor V A.; Santos, André A.C.; Veloso, Maria A.F.

2017-01-01

The spacer grids are structures present in nuclear fuel assembly of Pressurized Water Reactors (PWR). They play an important structural role and also assist in heat removal through the assembly by promoting increased turbulence of the flow. Understanding the flow dynamics downstream the spacer grid is paramount for fuel element design and analysis. This paper presents water flow velocity profiles measurements downstream a spacer grid in a 5 x 5 rod bundle test rig with the objective of highlighting important fluid dynamic behavior near the grid and supplying data for CFD simulation validation. These velocity profiles were obtained at two different heights downstream the spacer grid using a LDV (Laser Doppler Velocimetry) through the top of test rig. The turbulence intensities and patterns of the swirl and cross flow were evaluated. The tests were conducted for Reynolds numbers ranging from 1.8 x 10"4 to 5.4 x 10"4. This experimental research was carried out in thermo-hydraulics laboratory of Nuclear Technology Development Center – CDTN. Results show great repeatability and low uncertainties (< 1.24 %). Details of the flow field show how the mixture and turbulence induced by the spacer grid quickly decays downstream the spacer grid. It is shown that the developed methodology can supply high resolution low uncertainty results that can be used for validation of CFD simulations. (author)

A technical report on the evaluation of the integrity for the TIG welded spacer grid

International Nuclear Information System (INIS)

Song, Kee Nam; Yoo, Ho Sik; Lww, Chang Woo

1994-07-01

The spacer grid, which supports fuel rods, guide thimble and instrumentation tube, is classified into two types according to their strap material,.ie. inconel and zircaloy spacer grid. KOFA fuel of 14 x 14 and 17 x 17 type has seven and eight spacer grid respectively. Zircaloy spacer grid is assembled by straps whose cross points are welded by TIG welding method. This technical report provides to give some information about structure and function of the spacer grid and the basis and characteristic of the TIG welding method. A series of test which is conducted to evaluate the integrity of TIG welded zircaloy spacer grid and their results have been also studied. (Author) 18 refs., 23 figs., 3 tabs

A technical report on the evaluation of the integrity for the TIG welded spacer grid

Energy Technology Data Exchange (ETDEWEB)

Song, Kee Nam; Yoo, Ho Sik; Lww, Chang Woo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-07-01

The spacer grid, which supports fuel rods, guide thimble and instrumentation tube, is classified into two types according to their strap material,.ie. inconel and zircaloy spacer grid. KOFA fuel of 14 x 14 and 17 x 17 type has seven and eight spacer grid respectively. Zircaloy spacer grid is assembled by straps whose cross points are welded by TIG welding method. This technical report provides to give some information about structure and function of the spacer grid and the basis and characteristic of the TIG welding method. A series of test which is conducted to evaluate the integrity of TIG welded zircaloy spacer grid and their results have been also studied. (Author) 18 refs., 23 figs., 3 tabs.

CFD application to advanced design for high efficiency spacer grid

International Nuclear Information System (INIS)

Ikeda, Kazuo

2014-01-01

Highlights: • A new LDV was developed to investigate the local velocity in a rod bundle and inside a spacer grid. • The design information that utilizes for high efficiency spacer grid has been obtained. • CFD methodology that predicts flow field in a PWR fuel has been developed. • The high efficiency spacer grid was designed using the CFD methodology. - Abstract: Pressurized water reactor (PWR) fuels have been developed to meet the needs of the market. A spacer grid is a key component to improve thermal hydraulic performance of a PWR fuel assembly. Mixing structures (vanes) of a spacer grid promote coolant mixing and enhance heat removal from fuel rods. A larger mixing vane would improve mixing effect, which would increase the departure from nucleate boiling (DNB) benefit for fuel. However, the increased pressure loss at large mixing vanes would reduce the coolant flow at the mixed fuel core, which would reduce the DNB margin. The solution is to develop a spacer grid whose pressure loss is equal to or less than the current spacer grid and that has higher critical heat flux (CHF) performance. For this reason, a requirement of design tool for predicting the pressure loss and CHF performance of spacer grids has been increased. The author and co-workers have been worked for development of high efficiency spacer grid using Computational Fluid Dynamics (CFD) for nearly 20 years. A new laser Doppler velocimetry (LDV), which is miniaturized with fiber optics embedded in a fuel cladding, was developed to investigate the local velocity profile in a rod bundle and inside a spacer grid. The rod-embedded fiber LDV (rod LDV) can be inserted in an arbitrary grid cell instead of a fuel rod, and has the advantage of not disturbing the flow field since it is the same shape as a fuel rod. The probe volume of the rod LDV is small enough to measure spatial velocity profile in a rod gap and inside a spacer grid. According to benchmark experiments such as flow velocity

CFD application to advanced design for high efficiency spacer grid

Energy Technology Data Exchange (ETDEWEB)

Ikeda, Kazuo, E-mail: kazuo3_ikeda@ndc.mhi.co.jp

2014-11-15

Highlights: • A new LDV was developed to investigate the local velocity in a rod bundle and inside a spacer grid. • The design information that utilizes for high efficiency spacer grid has been obtained. • CFD methodology that predicts flow field in a PWR fuel has been developed. • The high efficiency spacer grid was designed using the CFD methodology. - Abstract: Pressurized water reactor (PWR) fuels have been developed to meet the needs of the market. A spacer grid is a key component to improve thermal hydraulic performance of a PWR fuel assembly. Mixing structures (vanes) of a spacer grid promote coolant mixing and enhance heat removal from fuel rods. A larger mixing vane would improve mixing effect, which would increase the departure from nucleate boiling (DNB) benefit for fuel. However, the increased pressure loss at large mixing vanes would reduce the coolant flow at the mixed fuel core, which would reduce the DNB margin. The solution is to develop a spacer grid whose pressure loss is equal to or less than the current spacer grid and that has higher critical heat flux (CHF) performance. For this reason, a requirement of design tool for predicting the pressure loss and CHF performance of spacer grids has been increased. The author and co-workers have been worked for development of high efficiency spacer grid using Computational Fluid Dynamics (CFD) for nearly 20 years. A new laser Doppler velocimetry (LDV), which is miniaturized with fiber optics embedded in a fuel cladding, was developed to investigate the local velocity profile in a rod bundle and inside a spacer grid. The rod-embedded fiber LDV (rod LDV) can be inserted in an arbitrary grid cell instead of a fuel rod, and has the advantage of not disturbing the flow field since it is the same shape as a fuel rod. The probe volume of the rod LDV is small enough to measure spatial velocity profile in a rod gap and inside a spacer grid. According to benchmark experiments such as flow velocity

Non linear fe analysis on the static buckling behavior of the spacer grid structures

International Nuclear Information System (INIS)

Song, K.N.; Yoon, K.H.

2001-01-01

In this study considered is the static buckling behavior of spacer grids in the fuel assembly, which are required to have a sufficient strength against an accident like earthquake. Special attention is given to the finite element modeling of the spot-welding and the constraints between the spacer strips assembled together: it is found that a proper treatment of the constraints is critical for accurate assessment of the buckling behavior including strain localization at the point of spot welding. The buckling strength of the 17 x 17 spacer grid, which is difficult to analyze due to a large number of degrees of freedom, is estimated from analysis for the smaller models 3 x 3, 5 x 5, 7 x 7, and 9 x 9 spacer grids. (authors)

Lower end fitting debris collector and end cap spacer grid

International Nuclear Information System (INIS)

Bryan, W.J.

1990-01-01

This patent describes a nuclear reactor having fuel assemblies including an upper end fitting and spaced nuclear fuel rod spacer grids for supporting and spacing a plurality of elongated nuclear fuel rods. Each includes a hollow active portion of nuclear fuel filled cladding intermediate the rod ends and tapering end cap of solid material with a circumferential groove on the rod end which first encounters reactor coolant flow, a lower end filtering debris collector and end cap spacer grid for capturing and retaining deleterious debris carried by reactor coolant before it enters the active region of a fuel assembly and creates fuel rod cladding damage

Nuclear reactor spring strip grid spacer

International Nuclear Information System (INIS)

Patterson, J.F.; Flora, B.S.

1980-01-01

An improved and novel grid spacer for maintaining the fuel rods of a nuclear reactor fuel assembly in substantially parallel array is described. The invention provides for spring strips to maintain the fuel elements in their desired orientation which have more positive alignment than previous types while allowing greater flexibility to counterbalance the effects of differential thermal expansion. (UK)

Nuclear reactor seismic fuel assembly grid

International Nuclear Information System (INIS)

Anthony, A.J.

1977-01-01

The strength of a nuclear reactor fuel assembly is enhanced by increasing the crush strength of the zircaloy spacer grids which locate and support the fuel elements in the fuel assembly. Increased resistance to deformation as a result of laterally directed forces is achieved by increasing the section modulus of the perimeter strip through bending the upper and lower edges thereof inwardly. The perimeter strip is further rigidized by forming, in the central portion thereof, dimples which extend inwardly with respect to the fuel assembly. The integrity of the spacer grid may also be enhanced by providing back-up arches for some or all of the integral fuel element locating springs and the strength of the fuel assembly may be further enhanced by providing, intermediate its ends, a steel seismic grid. 13 claims, 6 figures

Analyses of the Collapse Behavior of a Spacer Grid

International Nuclear Information System (INIS)

Jang, Myung-Geun; Na, Geum Ju; Jag, Yeon-Hui; Kim, Hee Cheol; Kim, Jong-Bong; Kim, Jaeyoug

2016-01-01

In order to investigate the protection capability of a spacer grid assembly for impact load, a hammer impact test has been carried out. The crush strength is measured in the hammer impact test. Song et al. carried out the experiment and finite element analysis for the hammer impact test for various weld line depth. Park et al. designed the spacer grid shape to get required crush strength via finite element analysis. Song et al. also optimized the spacer grid shape to maximize the crush strength, and carried out the finite element analysis for the hammer impact test considering the weld properties. Kim et al. carried out finite element analysis for various guide tube hole shape and compared the crush shape and crush strength. In this study, the effect of shape defect on the crush behavior in the hammer impact test is investigated. The spacer grid cannot be exactly the square. Therefore a lateral displacement (imperfection) is imposed to square spacer grid and then hammer impact is carried out. The effect of the lateral imperfection on the crush strength is investigated. The effect of the shape defect on the crushing behavior in the hammer impact test is investigated by finite element analysis. It is shown that the collapse become severe as the lateral imperfection displacement increases, especially when the imperfection is greater than or equal to 0.7 mm

Evaluation of spacer grid spring characteristics by means of physical tests and numerical simulation

Energy Technology Data Exchange (ETDEWEB)

Schettino, Carlos Frederico Mattos, E-mail: carlosschettino@inb.gov.br [Industrias Nucleares do Brasil (INB), Resende, RJ (Brazil)

2017-11-01

Among all fuel assemblies' components, the spacer grids play an important structural role during the energy generation process, mainly due for its primary functional requirement, that is, to provide fuel rod support. The present work aims to evaluate the spring characteristics of a specific spacer grid design used in a PWR fuel assembly type 16 x 16. These spring characteristics comprises the load versus deflection capability and its spring rate, which are very important, and also mandatory, to be correctly established in order to preclude spacer grid spring and fuel rod cladding fretting during operation, as well as prevent an excessive fuel rod buckling. This study includes physical tests and numerical simulation. The tests were performed on an adapted load cell mechanical device, using as a specimen a single strap of the spacer grid. Three numerical models were prepared using the Finite Element Method, with the support of the commercial code ANSYS. One model was built to validate the simulation according to the performed physical test, the others were built inserting a gradient of temperature (Beginning Of Life hot condition) and to evaluate the spacer grid spring characteristics in End Of Life condition. The obtained results from physical test and numerical model have shown a good agreement between them, therefore validating the simulation. The obtained results from numerical models make available information regarding the spacer grid design purpose, such as the behavior of the fuel rod cladding support during operation. Therewith, these evaluations could be useful to improve the spacer grid design. (author)

Evaluation of spacer grid spring characteristics by means of physical tests and numerical simulation

International Nuclear Information System (INIS)

Schettino, Carlos Frederico Mattos

2017-01-01

Among all fuel assemblies' components, the spacer grids play an important structural role during the energy generation process, mainly due for its primary functional requirement, that is, to provide fuel rod support. The present work aims to evaluate the spring characteristics of a specific spacer grid design used in a PWR fuel assembly type 16 x 16. These spring characteristics comprises the load versus deflection capability and its spring rate, which are very important, and also mandatory, to be correctly established in order to preclude spacer grid spring and fuel rod cladding fretting during operation, as well as prevent an excessive fuel rod buckling. This study includes physical tests and numerical simulation. The tests were performed on an adapted load cell mechanical device, using as a specimen a single strap of the spacer grid. Three numerical models were prepared using the Finite Element Method, with the support of the commercial code ANSYS. One model was built to validate the simulation according to the performed physical test, the others were built inserting a gradient of temperature (Beginning Of Life hot condition) and to evaluate the spacer grid spring characteristics in End Of Life condition. The obtained results from physical test and numerical model have shown a good agreement between them, therefore validating the simulation. The obtained results from numerical models make available information regarding the spacer grid design purpose, such as the behavior of the fuel rod cladding support during operation. Therewith, these evaluations could be useful to improve the spacer grid design. (author)

Numerical investigations on the effect of the axial interval between intensifying spacer grids on the critical heat flux value for fuel assemblies with non-uniform axial power distribution

International Nuclear Information System (INIS)

Kireeva, D.; Oleksyuk, D.

2015-01-01

In this paper a number of numerical studies on intensifying heat exchange conducted by NRC 'Kurchatov Institute' are presented. A standardised heat exchange intensifying spacer grid (UDRI) can be installed at any height along the fuel assembly (FA) heat-generating section. When installed at the bottom of a fuel assembly, the UDRI facilitates intensive coolant mixing; the UDRI mounted at the top of a FA provides better mixing and the enhancement in heat exchange. The application of the heat exchange intensifying spacer grids results in better flattening of the coolant parameters along the cross-section and higher critical heat flux ratio. The investigations were carried out by means of numerical code SC-INT using mesh generation that have been specially designed by NRC 'Kurchatov Institute' to perform calculations for fuel assemblies equipped with the intensifying spacer grids. The effect of the axial interval between UDRI grids on the critical heat flux value for two typical axial power shapes has been investigated. The derived optimal solutions for the positioning of intensifying grids are also presented

Effect of spacer grids on CHF in tube bundles

International Nuclear Information System (INIS)

Jayanti, Sreenivas; Valette, Michel

2004-01-01

Spacers grids are used to support tube bundles in steam generators and in nuclear reactor fuel assemblies. These grids interface with the flow and heat transfer in a number of ways and their effect has been studied by a number of researchers. It is known that generally they have a beneficial effect on critical heat flux (CHF) in typical nuclear reactor assemblies. However, the enhancement obtained depends on the geometric characteristics of the spacer grids as well as on the parameter range in terms of pressure, local mass velocity and quality. In the present study, the problem is approached in the context of a one-dimensional three-field model. Unlike in previous approaches, no specific modeling of the constitutive laws is made to account for spacer effects and only the geometric details such as the reduction in the cross-sectional area and the hydraulic diameter are included in the calculation which is otherwise the same as that for flow through a single tube. It is shown by comparison with literature data that this approach leads to satisfactory prediction of the thermal-hydraulic effects of spacers and that the beneficial effects of spacers on dry out can be manifested only when the entrainment rate is neither too high nor too low. Their effect on reducing the post-dry out wall temperature is also limited to certain cases. The present work has been performed as part of the EDF-CEA Neptune project also supported by the Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and FRAMATOME-ANP. NEPTUNE is a new set of two phase thermalhydraulic computer codes devoted to safety analysis of nuclear power plants. (author)

On the impact analysis of a PWR spacer grid

International Nuclear Information System (INIS)

Song, Kee Nam; Lee, S. H.

2012-01-01

A spacer grid, which is an interconnected array of slotted grid straps and is welded at the intersections to form an egg crate structure, is one of the most important structural components in a PWR fuel assembly. From a structural point of view, the spacer grid is required to have sufficient crush strength under lateral loads so that nuclear fuel rods are maintained in a cool able geometry, and that control rods can be inserted. The capacity of a spacer grid to resist lateral loads is usually characterized in terms of its crush strength, and it was reported that the lateral crush strength of the spacer grid is closely related with welding quality of the spacer grid. Microstructures in the weld zone, including a heat affected zone (HAZ), are different from that in a parent material. Consequently, the mechanical properties in the weld zone are different from those in the parent material to some extent. When a welded structure is loaded, the mechanical behavior of the welded structure might be different from the case of a structure with homogeneous mechanical properties. Nonetheless, mechanical properties in the welded structure have been neglected in many structural analyses of the spacer grid due to a lack of mechanical properties in the weld zone. When the weld zone is very narrow and the interfaces are not clear, it is difficult to take tensile test specimens in the weld zone. The reason for this is that the mechanical properties in the parent material are usually used in the structural analyses in the welded structure. As an aside, it has been recently determined that the ball indentation technique has the potential to be an excellent substitute for a standard tensile test, particularly in the case of small specimens or property gradient materials such as welds. In this study, to investigate the effect on the mechanical behavior of the spacer grid when using weld mechanical properties, strength analyses considering the weld mechanical properties recently obtained

Advanced KSNP fuel, plus7 : grid-to-rod fretting wear resistance of the plus7 spacer grids

International Nuclear Information System (INIS)

Kim, Kyu Tae; Kim, Yong Hwan; Jang, Young Ki; Choi, Joon Hyung

2003-01-01

Vibration-induced grid-to-rod fretting wear initiates at a certain critical gap correlated with a critical work rate. A critical gap between grid and rod forms due to in-reactor performance of fuel, thermal relaxation of grid spring and irradiation growth of grid strap, etc. A critical work rate may be generated by three vibration mechanisms proposed in this paper. Three vibration mechanisms have been derived with various fretting wear experience in commercial reactors as well as various out-of-pile hydraulic test results. The first active vibration mechanism is high turbulence-induced excessive fuel rod vibration with the combination of excessive grid-to-rod gap. The second active vibration mechanism is self-excited fuel assembly vibration in a low frequency range caused by hydraulically unbalanced mixing vanes of the spacer grid assembly. The third active vibration mechanism is self-excited spacer grid strap vibration in quite a high frequency range caused by some spacer grid designs. In this study, each vibration mechanism on the grid-to-rod fretting wear damage is discussed. On the other hand, the effects of various grid designs on the fretting wear damage in the commercial reactors are predicted using the long-term fretting wear test results. It is found that the larger grid-to-rod initial contact area generates the less fretting wear damage. Consequently the conformal spring of PLUS7 is superior to typical convex shaped spring with regard to fretting wear resistance since the former generates relatively larger contact area than the latter

CFD analysis on mixing effects of spacer grids with different dimples and sizes for advanced fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Yang, B.W.; Zhang, H.; Han, B.; Zha, Y.D.; Shan, J.Q. [Xi' an Jiaotong Univ. (China). School of Nuclear Science and Technology

2016-07-15

The thermal hydraulic characteristics of a mixing vane grid are largely dependent on the structure of key components, such as strip, spring, dimple, weld nugget, as well as the mixing vane configuration. In this paper, several types of spacer grids with different dimple shapes are modeled under subcooled boiling conditions. Prior to the application of CFD on the dimple shape analysis, the mixing effects of spacer grids were studied. After the dimple shape analysis, the side channel effect is discussed by comparing the simulation results of a 3 x 3 and a 5 x 5 spacer grid. The two phase flow CFD models in this study are validated through simple geometry showing that the calculated void fraction is in good agreement with the experimental data. The dimple comparison result shows that varying dimple structures can result in different temperatures, lateral velocities and void fraction distributions downstream of the spacer grids. Comparison of two sizes of spacer grids demonstrate that the side channel generates different flow distribution pattern in the center channel.

Pressure Drop Test of Hybrid Mixing Vane Spacer Grid

Energy Technology Data Exchange (ETDEWEB)

Oh, D. S.; Chang, S. K.; Kim, B. D.; Chun, S. Y.; Chun, T. H

2007-08-15

The pressure loss test has been accomplished in the test section containing 5x5 rod bundle with a length of 2 m including 3 spacer grids. The test has been performed for the 5 kinds of spacer grids to compare the pressure loss characteristics: 1. Plain spacer grid which has the same body of the Hybrid but without vane (Plain), 2. Hybrid Vane spacer grid (Hybrid), 3. Hybrid-SC spacer grid which is constructed with coined, chamfered strip and is fabricated by spot welding, 4. Hybrid-LC spacer grid which is constructed with coined, chamfered strip and is fabricated by line welding along intersection line, 5. Westinghouse spacer grid with split vane (Plus-7). The pressure loss coefficient of the Plain, Hybrid, Hybrid-SC, Hybrid-LC, and Plus-7 spacer grid is 0.93, 1.15, 1.02, 1.04, and 1.08, respectively.

Mixing vane grid spacer

International Nuclear Information System (INIS)

Patterson, J.F.; Galbraith, K.P.

1978-01-01

An improved mixing vane grid spacer having enhanced flow mixing capability by virtue of mixing vanes being positioned at welded intersecting joints of the spacer wherein each mixing vane has an opening or window formed therein substantially directly over the welded joint to provide improved flow mixing capability is described. Some of the vanes are slotted, depending on their particular location in the spacers. The intersecting joints are welded by initially providing consumable tabs at and within each window, which are consumed during the welding of the spacer joints

Procedure on the Impact Characteristic Test for the One-sided and Thru-grid Spacer Grid

International Nuclear Information System (INIS)

Yoon, Kyung Ho; Kim, Jae Yong; Lee, Kang Hee; Song, Kee Nam

2006-09-01

In order to perform the one-sided and the through-grid impact tests for a new developed spacer grid, the drop type impact test machines were established. The dynamic impact test is to get some basic data for accident analysis such as impact strength, stiffness, and coefficient of the restitution. Furthermore, these developed test methods and procedures will be qualified standard for increasing the reliability of the test results. Chapter 2 provides an introduction to the test facilities and instrumentations. Chapter 3 describes on how spacer grid and the single span fuel assembly specimen will be prepared. In addition to this, how to set up these testing machines. Chapter 4 illustrates detail test procedure on how to acquire impact signal of the two kinds of the specimen. Chapter 5 deals with signal processing and analysis for the test data. Finally, chapter 6 summarise the overall test procedure and the test method

Process development for the manufacturing of state-of-the-art spacer grids

Energy Technology Data Exchange (ETDEWEB)

Schebitz, Florian; Dietrich, Matthias [Advanced Nuclear Fuels GmbH, Karlstein (Germany)

2013-07-01

At the beginning it was questioned if 'time to market' is really important for the nuclear industry. The clear answer is YES. Even if the development times might be longer compared to projects in other industries it is still beneficial to use concurrent engineering. In the world wide network of manufacturing sites, Advanced Nuclear Fuels GmbH in Karlstein is quite often involved when the development of new processes is necessary. As ANF Karlstein is delivering products around the world the experience with different customer requirements supports an optimized solution in order to fulfill these principle requirements and to deliver state-of-the-art products like spacer grids. Continues feedback from process development already improves the first prototypes. In the meantime ANF Karlstein manufactured the components for both new fuel assembly designs which are introduced as a first set of Lead Fuel Assemblies. For the manufacturing of the next sets of spacer grids (for tests and next series of Lead Fuel Assemblies) the described processes will be used and further improved, so that an industrialized solution is available. (orig.)

Spacer grid for fuel elements

International Nuclear Information System (INIS)

Hensolt, T.; Huenner, M.; Rau, P.; Veca, A.

1978-01-01

The spacer grid for fuel elements of a gas-cooled fast breeder reactor (but also for PWRs and BWRs) consists of a lattice field with dodecagonal meshes. These meshes are formed by three each adjacent hexagons grouped arround a central axis. The pairs of legs extending into the dodecagon and being staggered by 120 0 are designed as knubs with inclined abutting surfaces for the fuel rods. By this means there is formed a three-point bearing for centering the fuel rods. The spacer grid mentioned above is rough-worked from a single disc- resp. plate-shaped body (unfinished piece). (DG) [de

Spacer grid for fuel elements

International Nuclear Information System (INIS)

Hensolt, T.; Huenner, M.; Rau, P.; Veca, A.

1980-01-01

The spacer grid for fuel elements of a gas-cooled fast breeder reactor (but also for PWRs and BWRs) consists of a lattice field with dodecagonal meshes. These meshes are formed by three each adjacent hexagons grouped arround a central axis. The pairs of legs extending into the dodecagon and being staggered by 120 are designed as knubs with inclined abutting surfaces for the fuel rods. By this means there is formed a three-point bearing for centering the fuel rods. The spacer grid mentioned above is rough-worked from a single disc- resp. plate-shaped body (unfinished piece). (orig.)

Hydraulic Design Criteria for Spacer Grids of Nuclear Fuel Element

International Nuclear Information System (INIS)

Juanico, Luis; Brasnarof, Daniel

2000-01-01

In this paper a hydraulic model for calculating the pressure drop on the CARA spacer grids is extended.This model is validated and feedback from experimental hydraulic test performed in a low pressure loop.The importance of the spacer grid geometric parameter (that is, its thickness and length, the number and kind of their fix spacer), developing hydraulic design criteria for spacer grid on fuel element

Comparative study of the contribution of various PWR spacer grid components to hydrodynamic and wall pressure characteristics

Energy Technology Data Exchange (ETDEWEB)

Bhattacharjee, Saptarshi, E-mail: saptarshi.bhattacharjee@outlook.com [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France); Ricciardi, Guillaume [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Viazzo, Stéphane [Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France)

2017-06-15

Highlights: • Complex geometry inside a PWR fuel assembly is simulated using simplified 3D models. • Structured meshes are generated as far as possible. • Fluctuating hydrodynamic and wall pressure field are analyzed using LES. • Comparative studies between square spacer grid, circular spacer grid and mixing vanes are presented. • Simulations are compared with experimental data. - Abstract: Flow-induced vibrations in a pressurized water reactor (PWR) core can cause fretting wear in fuel rods. These vibrations can compromise safety of a nuclear reactor. So, it is necessary to know the random fluctuating forces acting on the rods which cause these vibrations. In this paper, simplified 3D models like square spacer grid, circular spacer grid and symmetric mixing vanes have been used inside an annular pipe. Hydrodynamic and wall pressure characteristics are evaluated using large eddy simulations (LES). Structured meshes are generated as far as possible. Simulations are compared with an experiment. Results show that the grid and vanes have a combined effect: grid accelerates the flow whereas the vanes contribute to the swirl structures. Spectral analysis of the simulations illustrate vortex shedding phenomenon in the wake of spacer grids. This initial study opens up interesting perspectives towards improving the modeling strategy and understanding the complex phenomenon inside a PWR core.

Comparative study of the contribution of various PWR spacer grid components to hydrodynamic and wall pressure characteristics

International Nuclear Information System (INIS)

Bhattacharjee, Saptarshi; Ricciardi, Guillaume; Viazzo, Stéphane

2017-01-01

Highlights: • Complex geometry inside a PWR fuel assembly is simulated using simplified 3D models. • Structured meshes are generated as far as possible. • Fluctuating hydrodynamic and wall pressure field are analyzed using LES. • Comparative studies between square spacer grid, circular spacer grid and mixing vanes are presented. • Simulations are compared with experimental data. - Abstract: Flow-induced vibrations in a pressurized water reactor (PWR) core can cause fretting wear in fuel rods. These vibrations can compromise safety of a nuclear reactor. So, it is necessary to know the random fluctuating forces acting on the rods which cause these vibrations. In this paper, simplified 3D models like square spacer grid, circular spacer grid and symmetric mixing vanes have been used inside an annular pipe. Hydrodynamic and wall pressure characteristics are evaluated using large eddy simulations (LES). Structured meshes are generated as far as possible. Simulations are compared with an experiment. Results show that the grid and vanes have a combined effect: grid accelerates the flow whereas the vanes contribute to the swirl structures. Spectral analysis of the simulations illustrate vortex shedding phenomenon in the wake of spacer grids. This initial study opens up interesting perspectives towards improving the modeling strategy and understanding the complex phenomenon inside a PWR core.

Effect of spacer grid mixing vanes on coolant outlet temperature distribution

Energy Technology Data Exchange (ETDEWEB)

Raemae, Tommi; Lahtinen, Tuukka; Brandt, Tellervo; Toppila, Timo [Fortum Power and Heat, Fortum (Finland). Nuclear Competence Center

2012-08-15

In Loviisa VVER-440-type NPP the coolant outlet temperature of the hot subchannel is constantly monitored during the operation. According to the authority requirement the maximum subchannel outlet temperature must not exceed the saturation temperature. Coolant temperature distribution inside the fuel assembly is affected by the efficiency of the coolant mixing. In order to enhance the coolant mixing the fuel manufacturer is introducing the additional mixing vanes on the fuel bundle spacer grids. In the paper the effect of the different mixing vane modifications is studied with computational fluid dynamics (CFD) simulation. Goal of the modelling is to find vane modifications with which sufficient mixing is reached with acceptable increase in the spacer grid pressure loss. The results of the studies are discussed in the paper. (orig.)

Nuclear reactor spring strip grid spacer

International Nuclear Information System (INIS)

Patterson, J.F.; Flora, B.S.

1978-01-01

A bimetallic grid spacer is described comprising a grid structure of zircaloy formed by intersecting striplike members which define fuel element openings for receiving fuel elements and spring strips made of Inconel positioned within the grid structure for cooperating with the fuel elements to maintain them in their desired position. A plurality of these spring strips extend longitudinally between sides of the grid structure, being locked in position by the grid retaining strips. The fuel rods, which are disposed in the fuel openings formed in the grid structure, are positioned by means of the springs associated with the spring strips and a plurality of dimples which extend from the zircaloy grid structure into the openings. In one embodiment the strips are disposed in a plurality of arrays with those spring strip arrays situated in opposing diagonal quadrants of the grid structure extending in the same direction and adjacent spring strip arrays in each half of the spacer extending in relatively perpendicular directions. Other variations of the spring strip arrangements for a particular fuel design are disclosed herein

Side insertable spacer

International Nuclear Information System (INIS)

Patterson, J.F.; Ewing, R.H.

1992-01-01

This patent describes a spacer for restraining the fuel rods of a nuclear fuel assembly, the assembly being formed of a plurality of parallel, elongated fuel rods so arranged that the assembly is bounded by a polygon having an even number of sides, the rods being so arranged as to lie in a plurality of sets of parallel rows, the rows of each set being perpendicular to one of the sides of the polygon. It comprises a number of spacer combs equal to at least half the number of the sides of the polygon, the spacer combs being superposed on each other, each of the spacer combs comprising: a single base strip having a length equal to that of one of the sides of the polygon and grid strips equal in number to the spaces between rows in one of the sets, and at least a majority of the grid strips being of a length sufficient to extend substantially the full length of the rows; the grid strips being provided with spring members positioned to engage each of the rods; the grid strips being provided with spring members positioned to engage each of the rods; the grid strips being secured to and extending at right angles to the base strip; the grid strips of different combs being positioned at angles to each other, so as to occupy the spaces between rows in different sets

Fuel assembly spacer

International Nuclear Information System (INIS)

Shirakawa, Ken-etsu.

1988-01-01

Purpose: To reduce the pressure loss of coolants by fuel assembly spacers. Constitution: Spacers for supporting a fuel assembly are attached by means of a plurality of wires to an outer frame. The outer frame is made of shape memory alloy such that the wires are caused to slacken at normal temperature and the slacking of the wires is eliminated in excess of the transition temperature. Since the wires slacken at the normal temperature, fuel rods can be inserted easily. After the insertion of the fuel rods, when the entire portion or the outer frame is heated by water or gas at a predetermined temperature, the outer frame resumes its previously memorized shape to tighten the wires and, accordingly, the fuel rods can be supported firmly. In this way, since the fuel rods are inserted in the slacken state of the wires and, after the assembling, the outer frame resumes its memorized shape, the assembling work can be conducted efficiently. (Kamimura, M.)

A Study on Cell Size of Irradiated Spacer Grid for PWR Fuel

International Nuclear Information System (INIS)

Jin, Y. G.; Kim, G. S.; Ryu, W. S. and others

2014-01-01

The spacer grids supporting the fuel rods absorb vibration impacts due to the reactor coolant flow, and grid spring force decreases under irradiation. This reduction of contact force might cause grid-to-rod fretting wear. The fretting failure of the fuel rod is one of the recent significant issues in the nuclear industry from an economical as well as a safety concern. Thus, it is important to understand the characteristics of cell spring behavior and the change in size of grid cells for an irradiated spacer grid. In the present study, the dimensional measurement of a spacer grid was conducted to investigate the cell size of an irradiated spacer grid in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. To evaluate the fretting wear performance of an irradiated spacer grid, hot cell tests were carried out at IMEF of KAERI. Hot cell examinations include dimensional measurements for the irradiated spacer grid. The change of cell sizes was dependent on the direction of the spacer grids, leading to significant gap variations. It was found that the change in size of the cell springs due to irradiation-induced stress relaxation and creep during the fuel residency in the reactor core affect the contact behavior between the fuel rod and the cell spring

Preliminary analysis of the effect of the grid spacers on the reflood heat transfer

International Nuclear Information System (INIS)

Sugimoto, Jun; Murao, Yoshio

1982-02-01

The results are described about the preliminary analysis of the effect of the grid spacers on the heat transfer during reflood phase of a PWR LOCA. Experiments at JAERI and other facilities showed substantial heat transfer enhancement near the grid spacers. The heat transfer enhancement decreases with the distance from the grid spacers in the downstream region of the grid spacers. Several mechanisms are discussed about the heat transfer enhancement near the grid spacers. A model of a coalescence of the water droplets downstream the spacers is proposed based on the review of the experimental data. The heat transfer correlation for the saturated film boiling is utilized to quantify the heat transfer augmentation by the grid spacers. (author)

Nuclear reactor spring strip grid spacer

International Nuclear Information System (INIS)

Patterson, J.F.; Flora, B.S.

1980-01-01

An improved and novel grid spacer was developed for use in nuclear reactor fuel assemblies. It is comprised of a series of intersecting support strips and a peripheral support band attached to the ends of the support strips. Each of the openings into which the fuel element is inserted has a number of protruding dimples and springs extending in different directions. The dimples coact with the springs to secure the fuel rods in the openings. Compared with previous designs, this design gives more positive alignment of the support stips while allowing greater flexibility to counterbalance the effects of thermal expansion. The springs are arranged in alternating directions so that the reaction forces tend to counterbalance each other, which in turn minimizes the reaction loads on the supporting structure. (D.N.)

Spacer grid corner gusset

International Nuclear Information System (INIS)

Larson, J.G.

1984-01-01

There is provided a spacer grid for a bundle of longitudinally extending rods in spaced generally parallel relationship comprising spacing means for holding the rods in spaced generally parallel relationship; the spacing means includes at least one exterior grid strip circumscribing the bundle of rods along the periphery thereof; with at least one exterior grid strip having a first edge defining the boundary of the strip in one longitudinal direction and a second edge defining the boundary of the strip in the other longitudinal direction; with at least one exterior grid strip having at least one band formed therein parallel to the longitudinal direction; a plurality of corner gussets truncating each of a plurality of corners formed by at least one band and the first edge and the second edge

Nuclear reactor spacer assembly

International Nuclear Information System (INIS)

Anthony, A.J.; Groves, M.D.

1979-01-01

A fuel assembly for a nuclear reactor is disclosed wherein the fuel element receiving and supporting grid is comprised of a first metal, the guide tubes which pass through the grid assembly are comprised of a second metal and the grid is supported on the guide tubes by means of expanded sleeves located intermediate the grid and guide tubes. The fuel assembly is fabricated by inserting the sleeves, of initial outer diameter commensurate with the guide tube outer diameters, through the holes in the grid assembly provided for the guide tubes and thereafter expanding the sleeves radially outwardly along their entire length such that the guide tubes can subsequently be passed through the sleeves. The step of radial expansion, as a result of windows provided in the sleeves having dimensions commensurate with the geometry of the grid, mechanically captures the grid and simultaneously preloads the sleeve against the grid whereby relative motion between the grid and guide tube will be precluded

Air-water flooding in multirod channels: effects of spacer grids and blockages

International Nuclear Information System (INIS)

Cha, Jong Hee; Jun, Hyung Gil

1993-01-01

This paper presents the experimental results on flooding of countercurrent flow in vertical multirod channels, which consists of falling water film and upward air flow. In particular, the effects of spacer grids, with and without mixing vane, and of blockage in the multirod bundle on the behaviour of flooding were investigated. The 5 x 5 zircaloy tube bundle was used for the test section. The comparison of previous analytical models and empirical correlations with present data on flooding showed that the existing models and correlations predict much higher flooding curves. The spacer grid causes the lower flooding air flow rate to compare with the bare rod bundle. However, the mixing spacer grids need a higher flooding air flow rate for a constant liquid flow rate than the spacer grids without mixing vanes. The bundle containing blockages has the highest flooding air flow rate among the bundles with spacer grids and blokages. Empirical flooding correlations for the three types of test section have been made. (Author)

Detailed pressure drop measurements in single-and two-phase adiabatic air-water turbulent flows in realistic BWR fuel assembly geometry with spacer grids

International Nuclear Information System (INIS)

Caraghiaur, Diana; Frid, Wiktor; Tillmark, Nils

2004-01-01

In recent years, advance numerical simulation tools based on CFD methods have been increasingly used in various multi-phase flow applications. One of these is two-phase flow in fuel assemblies of Boiling Water Reactors. The important and often missing aspect of this development is validation of CFD codes against proper experimental data. The purpose of the current paper is to present detailed pressure measurements over a spacer grid in low pressure adiabatic single- and bubbly two-phase flow, which will be used to further develop a CFD code for BWR fuel bundle analysis. The experiments have been carried out in a n asymmetric 24-rod sub-bundle, representing one quarter of a Westinghouse SVEA-96 nuclear reactor fuel assembly. Single-phase flow measurements have been performed at superficial velocities between 0.90-4.50 m/s and in the two-phase flow, which was simulated by air-water mixture, measurements have been performed at void fractions ranging from 4 to 12% and liquid superficial velocity of 4.50 m/s. In order to increase the number of measuring points, five pressure taps were drilled in one of the rods, which was easily moved vertically by a traversing system, covering most of the points in axial direction. Any of the rods in the bundle could be substitute by the pressure sensing rod and the measurements were made for five pressure taps facing-angles. A detailed pressure distribution comparison between single- and two-phase flows for different sub-channel positions and different flow conditions was performed over one of the spacers. In addition, single-phase pressure drop measurements in the upper part of the test section comprising two spacer grids have been carried out. (author)

FIVPET Flow-Induced Vibration Test Report (1) - Candidate Spacer Grid Type I (Optimized H Type)

Energy Technology Data Exchange (ETDEWEB)

Lee, Kang Hee; Kang, Heung Seok; Yoon, Kyung Ho; Song, Kee Nam; Kim, Jae Yong

2006-03-15

The flow-induced vibration (FIV) test using a 5x5 partial fuel assembly was performed to evaluate mechanical/structural performance of the candidate spacer grid type I (Optimized H shape). From the measured vibration response of the test bundle and the flow parameters, design features of the spacer strap can be analyzed in the point of vibration and hydraulic aspect, and also compared with other spacer strap in simple comparative manner. Furthermore, the FIV test will contributes to understand behaviors of nuclear fuel in operating reactor. The FIV test results will be used to verify the theoretical model of fuel rod and assembly vibration. The aim of this report is to present the results of the FIV test of partial fuel assembly and to introduce the detailed test methodology and analysis procedure. In chapter 2, the overall configuration of test bundle and instrumented tube is remarked and chapter 3 will introduce the test facility (FIVPET) and test section. Chapter 4 deals with overall test condition and procedure, measurement and data acquisition devices, instrumentation equipment and calibration, and error analysis. Finally, test result of vibration and pressure fluctuation is presented and discussed in chapter 5.

Numerical investigation on practicability of reducing MCST by using grid spacer in a tight rod bundle

International Nuclear Information System (INIS)

Zhu, Xiaojing; Morooka, Shinichi; Oka, Yoshiaki

2014-01-01

Highlights: • Standard grid spacer design causes decreased heat transfer in a tight rod bundle. • Heat transfer is greatly enhanced by flow-enhancing features. • Swirling flow adversely affects the heat transfer downstream of grid spacer. • Enhanced heat transfer by existing grid spacer is limited in a short region. • Improved grid spacer can effectively reduce MCST. - Abstract: The numerical investigation was carried out to reveal the practicability of reducing the maximum cladding surface temperature (MCST) within the inner sub-channel of a tight, hexagon rod bundle using commercial CFD code STAR CCM+ 6.04. The special heat transfer and pressure drop characteristics caused by four existing grid spacer designs were discussed in detail by analyzing the effects of grid strap length, different flow enhancing features and different Reynolds numbers. It was found that the local heat transfer within the grid strap is greatly enhanced due to the raised flow velocity. Both the standard grid spacer and the grid spacer with split-vanes cause decreased heat transfer in the downstream region. The friction drag is very influential in the tight rod bundle and can eliminate the positive effect of flow blockage on the heat transfer performance. The grid spacer with flow blockage discs induces relatively good heat transfer performance and higher pressure drop within sub-channels, indicating a tradeoff between the heat transfer augmentation and the pressure drop. The combination of multiple existing grid spacers can reduce the MCST to a certain level, but the corresponding disadvantages cannot be ignored. The improved grid spacer design was proposed based on the overall considerations of heat transfer and pressure drop characteristics and has been proved more suitable to widely reduce MCST for SCWR than any other grid spacer designs involved in present study

Experimental investigation of turbulent flow through spacer grids in fuel rod bundles

International Nuclear Information System (INIS)

Caraghiaur, Diana; Anglart, Henryk; Frid, Wiktor

2009-01-01

This paper contains experimental data of pressure, velocity and turbulence intensity in a 24-rod fuel bundle with spacer grids. Detailed pressure measurements inside the spacer grid have been obtained by use of a sliding pressure-sensing rod. Laser Doppler Velocimetry technique was used to measure the local axial velocity and its fluctuating component upstream and downstream of the spacer grid in sub-channels with different blockage ratios. The measurements show a changing pattern in function of radial position in the cross-section of the fuel bundle. For sub-channels close to the box wall, the turbulence intensity suddenly increases just downstream of the spacer and then gradually decays. In inner sub-channels, however, the turbulence intensity downstream of the spacer decreases below its upstream value and then gradually increases until it reaches the maximum value at approximately two spacer heights. The present study reveals that spacer effects, such as local pressure distribution and turbulence intensity enhancement, not only depend exclusively on the local geometry details, but also on the location in the cross-section of the rod bundle.

Experimental investigation of turbulent flow through spacer grids in fuel rod bundles

Energy Technology Data Exchange (ETDEWEB)

Caraghiaur, Diana [Royal Institute of Technology, Division of Nuclear Reactor Technology, Department of Physics, School of Engineering Sciences, AlbaNova University Center, SE-106 91 Stockholm (Sweden)], E-mail: dianac@kth.se; Anglart, Henryk [Royal Institute of Technology, Division of Nuclear Reactor Technology, Department of Physics, School of Engineering Sciences, AlbaNova University Center, SE-106 91 Stockholm (Sweden); Frid, Wiktor [Swedish Radiation Safety Authority, Reactor Technology and Structural Integrity, SE-171 16 Stockholm (Sweden)

2009-10-15

This paper contains experimental data of pressure, velocity and turbulence intensity in a 24-rod fuel bundle with spacer grids. Detailed pressure measurements inside the spacer grid have been obtained by use of a sliding pressure-sensing rod. Laser Doppler Velocimetry technique was used to measure the local axial velocity and its fluctuating component upstream and downstream of the spacer grid in sub-channels with different blockage ratios. The measurements show a changing pattern in function of radial position in the cross-section of the fuel bundle. For sub-channels close to the box wall, the turbulence intensity suddenly increases just downstream of the spacer and then gradually decays. In inner sub-channels, however, the turbulence intensity downstream of the spacer decreases below its upstream value and then gradually increases until it reaches the maximum value at approximately two spacer heights. The present study reveals that spacer effects, such as local pressure distribution and turbulence intensity enhancement, not only depend exclusively on the local geometry details, but also on the location in the cross-section of the rod bundle.

Technology of electron beam welding for Zr-4 alloy spacer grid

International Nuclear Information System (INIS)

Pei Qiusheng; Wu Xueyi; Yang Qishun

1989-10-01

The welding technology for Zr-4 alloy spacer grid by using vacuum electron beam was studied. Through a series of welding technological experiments, metallographic examinations of seam structure and detecting tests for welding defect by X-ray defectoscopy, a good welding technology was selected to meet the requirements. The experimental results indicated that the Zr-4 alloy spacer grid welded by vacuum electron beam welding is feasible

A simulation methodology of spacer grid residual spring deflection for predictive and interpretative purposes

International Nuclear Information System (INIS)

Kim, K. T.; Kim, H. K.; Yoon, K. H.

1994-01-01

The in-reactor fuel rod support conditions against the fretting wear-induced damage can be evaluated by spacer grid residual spring deflection. In order to predict the spacer grid residual spring deflection as a function of burnup for various spring designs, a simulation methodology of spacer grid residual spring deflection has been developed and implemented in the GRIDFORCE program. The simulation methodology takes into account cladding creep rate, initial spring deflection, initial spring force, and spring force relaxation rate as the key parameters affecting the residual spring deflection. The simulation methodology developed in this study can be utilized as an effective tool in evaluating the capability of a newly designed spacer grid spring to prevent the fretting wear-induced damage

Effect of Weld Properties on the Crush Strength of the PWR Spacer Grid

Directory of Open Access Journals (Sweden)

Kee-nam Song

2012-01-01

Full Text Available Mechanical properties in a weld zone are different from those in the base material because of different microstructures. A spacer grid in PWR fuel is a structural component with an interconnected and welded array of slotted grid straps. Previous research on the strength analyses of the spacer grid was performed using base material properties owing to a lack of mechanical properties in the weld zone. In this study, based on the mechanical properties in the weld zone of the spacer grid recently obtained by an instrumented indentation technique, the strength analyses considering the mechanical properties in the weld zone were performed, and the analysis results were compared with the previous research.

Modal properties of the flexural vibrating package of rods linked by spacer grids

Directory of Open Access Journals (Sweden)

Zeman V.

2011-06-01

Full Text Available The paper deals with the modelling and modal analysis of the large package of identical parallel rods linked by transverse springs (spacer grids placed on several level spacings. The rod discretization by finite element method is based on Rayleigh beam theory. For the cyclic and central symmetric package of rods (such as fuel rods in nuclear fuel assembly the system decomposition on the identical revolved rod segments was applied. A modal synthesis method with condensation is used for modelling of the whole system. The presented method is the first step for modelling the nuclear fuel assembly vibration caused by excitation determined by the support plate motion of the reactor core.

Numerical simulation of bubble motion about a grid spacer in a rod bundle

International Nuclear Information System (INIS)

Zhang, Zheng; Hosokawa, Shigeo; Hayashi, Kosuke; Tomiyama, Akio

2009-01-01

Numerical simulations based on a three-dimensional two-way bubble tracking method are carried out to predict bubble motions in a square duct with an obstacle and in a two-by-three rod bundle with a grid spacer. Comparisons between measured and predicted bubble motions demonstrate that the two-way bubble tracking method gives good predictions for trajectories of small bubbles in the upstream side of the grid spacer in the rod bundle geometry. The predicted bubble trajectories clearly show that bubbles are apt to migrate toward the rod surface in the vicinity of the bottom of the grid spacer. Analysis of forces acting on the bubbles confirms that pressure gradient force induced by the presence of the spacer is the main cause of the bubble lateral migration toward the rod surface. Motions of steam bubbles at a nominal operating condition of a pressurized water reactor (PWR) are also predicted by using the bubble tracking method, which indicates that steam bubbles also migrate toward the rod surface at the upstream side of the spacer due to the spacer-induced pressure gradient force. (author)

Geometric effects of spacer grid in an annulus flow channel during reflooding period

International Nuclear Information System (INIS)

Cho, S.; Chun, S. Y.; Kim, B. D.; Park, J. K.; Yun, Y. J.; Baek, W. P.

2004-01-01

A number of studies on the reflooding phase were actively carried out from the early 70's due to its importance for the safety of the nuclear reactor. (Martini et al., 1973; Henry, 1974; Chung, 1978;) However, few studies have presented the spacer grid effect during the reflooding period. Since the grid is an obstruction in the flow passage, it causes an increased pressure drop due to form and skin friction losses. On the other hand, the spacer grid tends to increase the local wall heat transfer. The present work has been performed in a vertical annulus flow channel with various flow conditions. The objective of this paper is to evaluate the effects of a swirl-vane spacer grid on the rewetting phenomena during the reflooding phase

Tube spacer grid for a heat-exchanger tube bundle

International Nuclear Information System (INIS)

Scheidl, H.

1976-01-01

A tube spacer grid for a heat-exchanger tube bundle is formed by an annular grid frame having a groove formed in its inner surface in which the interspaced grid bars have their ends positioned and held in interspaced relationship by short sections of tubes passed through holes axially formed in the grid frame so that the tubes are positioned between the ends of the grid bars in the grooves. The tube sections may be cut from the same tubes used to form the tube bundle. 5 claims, 3 drawing figures

Bimetallic spacer means for a nuclear fuel assembly

International Nuclear Information System (INIS)

Anthony, A.J.

1981-01-01

A bimetallic spacer means designed to be cooperatively associated with a nuclear fuel assembly and operative to resist the occurrence of in-reactor bowing of the nuclear fuel assembly. The subject bimetallic spacer means in accord with one embodiment of the invention includes a member formed, at least principally, of Zircaloy to which are attached a plurality of stainless steel strips. The latter stainless steel strips are located on the external surface of the Zircaloy member and with the major axis of each of the plurality of stainless steel strips extending substantially perpendicular to the major axis of the Zircaloy member. In accord with another embodiment of the invention, the subject bimetallic spacer means includes a member formed at least principally of Zircaloy to which a plurality of stainless steel strips are attached so as to be positioned thereon externally thereof and with the major axis of each of the plurality of stainless steel strips extending substantially parallel to the major axis of the Zircaloy member. In accord with a further embodiment of the invention, the stainless steel strips are attached to preselected members, each embodying at least a cladding of Zircaloy, which are located in the rows of fuel rods that define the perimeter of the fuel matrix of the nuclear fuel assembly. In each of the embodiments, the stainless steel strips during power production expand outwardly to a greater extent than do the members to which the stainless steel strips are attached, thereby forming stiff springs which abut against like bimetallic spacer means with which the other nuclear fuel assemblies are provided in a given nuclear reactor core to thus prevent the occurrence of in-reactor bowing of the nuclear fuel assemblies. Namely, the stainless steel strips expand laterally relative to the fuel assembly and thereby occupy the space adjacent to the external surface of the fuel assembly

Measurement and model development of the droplet diameter in rod bundles with spacer grids in the reactor core

Energy Technology Data Exchange (ETDEWEB)

No, Hee Cheon; Lee, Eo Hwak; Yoo, Seung Hun; Jin, Hyung Gon; Kim, In Hun [KAIST, Daejeon (Korea, Republic of)

2010-05-15

To understand and to predict the heat transfer between superheated steam and droplets properly during reflood phase of LBLOCA of APR1400, it is very important to measure broken droplet sizes by spacer grids. A study, therefore, has been performed to investigate droplet size in rod bundles with spacer grids and to develop a spacer grid droplet size model for safety analysis codes. Experiments were conducted with liquid droplets (SMD of 300{approx}700 {mu}m) impacting on various spacer grids at air superficial velocity of 10 and 20 m/s based on FLECHT SEASET. The test channel and the grids were heated to 150 .deg. C to prevent the formation of liquid film during tests. The spacer grids were designed refer to the Korean fuel rod bundles (Korean Standard Fuel, Plus 7) of APR1400 with various blockage area ratio and grid geometries (strap thickness, mixing vane) and about 15,000 droplets were measured at upstream and downstream of the grids in 16 tests. As a result, the measurement of broken droplet size by spacer grids with photography method is presented and the droplet size model related to spacer grids as a function of blockage area ratio is suggested in this report

Enhancement of nuclear heat transfer in a typical pressurized water reactor by new spacer grids

International Nuclear Information System (INIS)

Nazifi, M.; Nematollahi, M.

2007-01-01

The fuel element geometry typically used in nuclear reactor is rod bundle whose rod-to-rod clearance is maintained by grid spacer. The heat generated in the rod by nuclear reaction is removed by coolant, usually in turbulent flow. The coolant moves axially through the subchannels. Fuel spacer grid affects the coolant flow distribution in a fuel rod bundle, and so spacer geometry has a strong influence on a bundle's thermal-hydraulic characteristics such as critical heat flux and pressure drop. An understanding of the detailed structure of the turbulent flow and heat transfer in the rod bundle, used especially as nuclear fuel elements, is of major interest to the nuclear power industry for their safe and reliable operation. The flow mixing devices on grid spacer would enhance the mixing rate between sub-channels and promote the turbulence in subchannel. The present study evaluates the effects of mixing vane shape on flow structure and heat transfer downstream of mixing vane in a sub-channel of fuel assembly, by obtaining velocity and pressure fields, turbulent intensity, flow mixing factors, heat transfer coefficient and friction factor using three-dimensional RANS analysis. Six new shapes mixing vane designed by the authors, are simulated numerically to evaluate the performance in enhancing the heat transfer, in comparison with commercialized split vane. Standard K-epsilon model are used as a turbulence closure model and periodic and symmetry condition are set as boundary conditions. The capability of the model to predict the coolant flow distribution inside rod bundles is shown and discussed on the base of comparison with experimental data for a variety of geometrical and Reynolds number conditions. It is conformed that the turbulence in the sub-channel was significantly promoted by spacer and mixing devices but rapidly decreased to a fully developed level approximately 10 time of hydraulic diameter downstream of the top of spacer. Ring type mixer showed a high

Reentrainment of droplet from grid spacer in mist flow portion of LOCA reflood of PWR

International Nuclear Information System (INIS)

Lee, S.L.; Cho, S.K.; Sheen, H.J.

1983-01-01

An investigation is made on the influence of a quenched grid spacer on the greatly enhanced heat transfer from heated fuel rods during the mist flow phase of emergency reflood of loss of coolant accident (LOCA) of a pressurized water reactor (PWR). The situation for the case of a dry grid spacer before its quenching has not been covered in this study. The experimental technique used is a relatively simple optical scheme which combines the reference-mode laser-Doppler anemometry making use of the scattering of a light beam from a droplet. The results reveal that the large droplets in the mist flow, which are intercepted by the grid spacer, are responsible for the creation of a large number of smaller droplets. These small droplets, due to their large surface area to mass ratios, can serve as superb evaporative cooling agents to heat transfer downstream of the grid spacer

An impact test system design and its applications to dynamic buckling of a spacer grid assembly

Energy Technology Data Exchange (ETDEWEB)

Liu, Sheng, E-mail: liusheng_05@126.com; Fan, Chenguang; Yang, Yiren

2016-11-15

This study is aimed at investigating the dynamic buckling load, dynamic stiffness, damping and buckling characteristics of the spacer grid assembly (SGA). A pendulum impact test system is designed to experiment the buckling of SGAs. Three criterions are discussed and compared to determine the buckling loads of SGAs: B-R criterion, energy criterion and extreme value criterion. Two approaches are applied to calculate the dynamic stiffness of SGAs: One method is natural period method based on the hypothesis of harmonic motion of the pendulum whose period is approximated because of the passivation and tailing of the impact force time history; and the other is energy method based on the conservation of mechanical energy. The equivalent viscous damping is defined as the resultant cause of dissipation and is obtained by the energy principle. The impact force time history loses its approximate symmetry after buckling occurs. The impact force and displacement reach their maxima almost at the same time at pre-buckling states but not post-buckling states. Vertical straps in SGA are found to be transversely shared by horizontal straps at the buckling position. The buckling of SGA results from the lack of strength of complete structure; and the strength of material has no effects on the buckling.

Vibration characteristics of a PWR fuel rod supported by optimized H type spacer grids

International Nuclear Information System (INIS)

Choi, M. H.; Kang, H. S.; Yoon, K. H.; Kim, H. K.; Song, K. N.

2002-01-01

The spacer grids are one of the main structural components in the fuel assembly, which supports and protects the fuel rods from the external loads by seismic and coolant flow. In this study, a modal test and a FE vibration analysis using ABAQUS are performed on a PWR dummy fuel rod of 3.847 m which is continuously supported by eight Optimized H type spacer grids. The experimental results agree with previous works that the natural frequencies decrease, while the amplitudes increase, with the increase of the excitation force. The force levels showing the maximum displacement of 0.2 mm are in the range from 0.2 N to 0.3 N, and at the same force range the fundamental frequencies are measured around 42.0 Hz, at which the relatively big displacements are observed at the 7th span. The results from the modal tests and the FE analyses are compared by both Modal Assurance Criteria (MAC) values and mode shapes. The MAC values at 2nd, 4th, and 7th mode are below 50%. It is believed that the reason of the low MACs at those modes is that the vibration amplitudes of the modes are more distorted by the excitation force than those of the other modes

Effects of grid spacer with mixing vane on entrainments and depositions in two-phase annular flows

Directory of Open Access Journals (Sweden)

Akimaro Kawahara

2015-06-01

Full Text Available The effects of mixing vanes (MVs attached to a grid spacer on the characteristics of air–water annular flows were experimentally investigated. To know the effects, a grid spacer with or without MV was inserted in a vertical circular pipe of 16-mm internal diameter. For three cases (i.e., no spacer, spacer without MV, and spacer with MV, the liquid film thickness, liquid entrainment fraction, and deposition rate were measured by the constant current method, single liquid film extraction method, and double liquid film extraction method, respectively. The MVs significantly promote the re-deposition of liquid droplets in the gas core flow into the liquid film on the channel walls. The deposition mass transfer coefficient is three times higher for the spacer with MV than for the spacer without MV, even for cases 0.3-m downstream from the spacer. The liquid film thickness becomes thicker upstream and downstream for the spacer with MV, compared with the thickness for the spacer without MV and for the case with no spacer.

Preliminary Investigation on Turbulent Flow in Tight-lattice Rod Bundle with Twist-mixing Vane Spacer Grid

International Nuclear Information System (INIS)

Lee, Chiyoung; Kwack, Youngkyun; Park, Juyong; Shin, Changhwan; In, Wangkee

2013-01-01

Our research group has investigated the effect of P/D difference on the behavior of turbulent rod bundle flow without the mixing vane spacer grid, using PIV (Particle Image Velocimetry) and MIR (Matching Index of Refraction) techniques for tight lattice fuel rod bundle application. In this work, using the tight-lattice rod bundle with a twist-mixing vane spacer grid, the turbulent rod bundle flow is preliminarily examined to validate the PIV measurement and CFD (Computational Fluid Dynamics) simulation. The turbulent flow in the tight-lattice rod bundle with a twist-mixing vane spacer grid was preliminarily examined to validate the PIV measurement and CFD simulation. Both were in agreement with each other within a reasonable degree of accuracy. Using PIV measurement and CFD simulation tested in this work, the detailed investigations on the behavior of turbulent rod bundle flow with the twist-mixing vane spacer grid will be performed at various conditions, and reported in the near future

CFD analysis of pressure drop across grid spacers in rod bundles compared to correlations and heavy liquid metal experimental data

Energy Technology Data Exchange (ETDEWEB)

Batta, A., E-mail: batta@kit.edu; Class, A.G., E-mail: class@kit.edu

2017-02-15

Early studies of the flow in rod bundles with spacer grids suggest that the pressure drop can be decomposed in contributions due to flow area variations by spacer grids and frictional losses along the rods. For these shape and frictional losses simple correlations based on theoretical and experimental data have been proposed. In the OECD benchmark study LACANES it was observed that correlations could well describe the flow behavior of the heavy liquid metal loop including a rod bundle with the exception of the core region, where different experts chose different pressure-loss correlations for the losses due to spacer grids. Here, RANS–CFD simulations provided very good data compared to the experimental data. It was observed that the most commonly applied Rehme correlation underestimated the shape losses. The available correlations relate the pressure drop across a grid spacer to the relative plugging of the spacer i.e. solidity e{sub max}. More sophisticated correlations distinct between spacer grids with round or sharp leading edge shape. The purpose of this study is to (i) show that CFD is suitable to predict pressure drop across spacer grids and (ii) to access the generality of pressure drop correlations. By verification and validation of CFD results against experimental data obtained in KALLA we show (i). The generality (ii) is challenged by considering three cases which yield identical pressure drop in the correlations. First we test the effect of surface roughness, a parameter not present in the correlations. Here we compare a simulation assuming a typical surface roughness representing the experimental situation to a perfectly smooth spacer surface. Second we reverse the flow direction for the spacer grid employed in the experiments which is asymmetric. The flow direction reversal is chosen for convenience, since an asymmetric spacer grid with given blockage ratio, may result in different flow situations depending on flow direction. Obviously blockage

Interaction between zircaloy tube and inconel spacer grid at high temperature

International Nuclear Information System (INIS)

Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi; Furuta, Teruo

1990-09-01

In order to investigate the interaction between fuel cladding and spacer grid of the pressurized water reactor during a severe accident, isothermal reaction tests were performed at the temperature range from 1248 to 1673K. A specimen consisted of a short Zircaloy-4 cladding tube and a piece of spacer grid of Inconel-718. In the tests in an argon atmosphere, eutectic reaction between Zircaloy and Inconel was observed at the contact points at 1248K. Rapid reaction was observed at higher test temperatures. For example, in the test at 1373K for 300s, Zircaloy reacted with Inconel over the entire thickness (0.62mm) of the tube in the vicinity of the contact point. In the present tests, Zircaloy which has higher melting point than Inconel was dissolved preferentially due to eutectic formation. In the tests in an oxygen atmosphere, no eutectic reaction was observed at temperatures below 1437K. A trace of interaction was found at the contact point of specimen heated at 1573 and 1623K. However, decrease in Zircaloy thickness was not measured. The possibility of eutectic reaction between Zircaloy cladding and Inconel spacer grid seems to be quite limited when sufficient oxygen is supplied. (author)

Experimental benchmark data for PWR rod bundle with spacer-grids

International Nuclear Information System (INIS)

Dominguez-Ontiveros, Elvis E.; Hassan, Yassin A.; Conner, Michael E.; Karoutas, Zeses

2012-01-01

In numerical simulations of fuel rod bundle flow fields, the unsteady Navier–Stokes equations have to be solved in order to determine the time (phase) dependent characteristics of the flow. In order to validate the simulations results, detailed comparison with experimental data must be done. Experiments investigating complex flows in rod bundles with spacer grids that have mixing devices (such as flow mixing vanes) have mostly been performed using single-point measurements. In order to obtain more details and insight on the discrepancies between experimental and numerical data as well as to obtain a global understanding of the causes of these discrepancies, comparisons of the distributions of complete phase-averaged velocity and turbulence fields for various locations near spacer-grids should be performed. The experimental technique Particle Image Velocimetry (PIV) is capable of providing such benchmark data. This paper describes an experimental database obtained using two-dimensional Time Resolved Particle Image Velocimetry (TR-PIV) measurements within a 5 × 5 PWR rod bundle with spacer-grids that have flow mixing vanes. One of the unique characteristic of this set-up is the use of the Matched Index of Refraction technique employed in this investigation to allow complete optical access to the rod bundle. This unique feature allows flow visualization and measurement within the bundle without rod obstruction. This approach also allows the use of high temporal and spatial non-intrusive dynamic measurement techniques namely TR-PIV to investigate the flow evolution below and immediately above the spacer. The experimental data presented in this paper includes explanation of the various cases tested such as test rig dimensions, measurement zones, the test equipment and the boundary conditions in order to provide appropriate data for comparison with Computational Fluid Dynamics (CFD) simulations. Turbulence parameters of the obtained data are presented in order to gain

Experimental investigations of the post-CHF heat transfer of R-134a flow-boiling in an annulus with spacer grids

International Nuclear Information System (INIS)

Lee, Kwi Lim; Chang, Soon Heung

2009-01-01

An experimental study was performed in the post-CHF condition using R-134a to investigate the effect of spacer grids on post-CHF heat transfer in an annulus channel. The experiments were conducted under the outlet pressures of 1.1 - 2.0 MPa, the mass fluxes of 100 - 400 kg/m 2 s and the inlet temperatures of 25 - 51degC. About 300 data of post-CHF data were obtained in the annular geometry without spacer grids and compared with several post-CHF correlations. The results showed the large prediction uncertainty mainly caused by the cold wall effect, so the empirical correlation for an annulus geometry without spacer grids was developed with the present experimental results. The heat transfer coefficient was calculated based on the heater rod temperature and the saturated vapor property. The average and root-mean-square(RMS) errors of the predictions were 0.17 % and 3.4 %, respectively. The experiments related to the spacer grid effects were performed with an I-type spacer grid and split-swirl mixing vane (with blockage-area ratios of 4.0 and 5.8 %). The spacer grid and mixing vane test results showed the enhancing effect on the heat transfer at the downstream location of the spacers. The experimental results from the spilt-swirl-type grid tests were more effective than the I-type grid tests. This was attributed to enhance the turbulence and increase the heat transfer caused by the mixing vane. (author)

Measurement of droplet dynamics across grid spacer in mist cooling of subchannel of PWR

International Nuclear Information System (INIS)

Lee, S.L.; Sheen, H.J.; Cho, S.K.; Issapour, I.

1984-01-01

An experiment was conducted of the dynamics and heat transfer of a droplet-vapor mist flow across a test grid spacer in a flow channel of 2 x 2 electrically heated simulation fuel rods. Embedded thermocouples were used to measure the rod cladding temperature and an unshielded Chromel-Alumel thermocouple was transversed in the center of the subchannel to measure the temperature of the water and steam coolant phases at various axial locations. Thermocouples were also embedded in the test grid spacer. Optical measurements of the size and velocity distributions of droplets and the velocity distribution of the superheated steam were made by special laser-Doppler anemometry techniques through quartz glass windows immediately upstream and downstream of the test grid spacer. Experiments over a range of steam and injected water flow rates and rod heat flux have been performed and some representative results and discussions are presented

The effect of spacer grid critical component on pressure drop under both single and two phase flow conditions

International Nuclear Information System (INIS)

Han, B.; Yang, B.W.; Zhang, H.; Mao, H.; Zha, Y.

2016-01-01

As pressure drop is one of the most critical thermal hydraulic parameters for spacer grids the accurate estimation of it is the key to the design and development of spacer grids. Most of the available correlations for pressure drop do not contain any real geometrical parameters that characterize the grid effect. The main functions for spacer grid are structural support and flow mixing. Once the boundary sublayer near the rod bundle is disturbed, the liquid forms swirls or flow separation that affect pressure drop. However, under two phase flow conditions, due to the existence of steam bubble, the complexity for spacer grid are multiplied and pressure drop calculation becomes much more challenging. The influence of the dimple location, distance of mixing vane to the nearest strip, and the effect of inter-subchannel mixing among neighboring subchannels on pressure drop and downstream flow fields are analyzed in this paper. Based on this study, more detailed space grid geometry parameters are recommended for adding into the correlation when predicting pressure drop.

Analysis of the optimized H type grid spring by a characterization test and the finite element method under the in-grid boundary

International Nuclear Information System (INIS)

Yoon, Kyung Ho; Lee, Kang Hee; Kang, Heung Seok; Song, Kee Nam

2006-01-01

Characterization tests (load vs. displacement curve) are conducted for the springs of Zirconium alloy spacer grids for an advanced LWR fuel assembly. Twofold testing is employed: strap-based and assembly-based tests. The assembly-based test satisfies the in situ boundary conditions of the spring within the grid assembly. The aim of the characterization test via the aforementioned two methods is to establish an appropriate assembly-based test method that fulfills the actual boundary conditions. A characterization test under the spacer grid assembly boundary condition is also conducted to investigate the actual behavior of the spring in the core. The stiffness of the characteristic curve is smaller than that of the strap-wised boundary condition. This phenomenon may cause the strap slit condition. A spacer grid consists of horizontal and vertical straps. The strap slit positions are differentiated from each other. They affords examination of the variation of the external load distribution in the grid spring. Localized regions of high stress and their values are analyzed, as they may be affected by the spring shape. Through a comparison of the results of the test and FE analysis, it is concluded that the present assembly-based analysis model and procedure are reasonably well conducted and can be used for spring characterization in the core. Guidelines for improving the mechanical integrity of the spring are also discussed

Nuclear fuel assembly seismic amplitude limiter

International Nuclear Information System (INIS)

Anthony, A.J.

1977-01-01

The ability of a nuclear reactor to withstand high seismic loading is enhanced by including, on each fuel assembly, at least one seismic grid which reduces the magnitude of the possible lateral deflection of the individual fuel elements and the entire fuel assembly. The reduction in possible deflection minimizes the possibility of impact of the spacer grids of one fuel assembly on those of an adjacent fuel assembly and reduces the magnitude of forces associated with any such impact thereby minimizing the possibility of fuel assembly damage as a result of high seismic loading. The seismic grid is mounted from the fuel assembly guide tubes, has greater external dimensions when compared to the fuel assembly spacer grids and normally does not support or otherwise contact the fuel elements. The reduction in possible deflection is achieved through reduction of the clearance between adjacent fuel assemblies made possible by the use in the seismic grid of a high strength material characterized by favorable thermal expansion characteristics and minimal irradiation induced expansion

A Study on Abrasive Wear Behavior of Spacer Grid Materials for Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Park, J. M.; Kim, J. H. [Chungnam National University, Daejeon (Korea, Republic of); Park, J. K.; Jeon, K. L. [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

2010-10-15

Spacer grid is one of the key components of a light water reactor (LWR) fuel assembly. The most important function of it is to hold the fuel rods to maintain the distance between the fuel rods inside a fuel assembly. At the reactor core in operating power plants, a fretting damage has been frequently reported between a nuclear fuel rod and its supporting spring/dimple of the fuel assemblies. This is due to a flow induced vibration (FIV), Which results from the primary coolant that rapidly passes around the fuel rod to remove the excess heat generated by the nuclear reaction. Fretting damage is generally caused by fretting wear, which includes various wear mechanisms such as an oxidative, adhesive, abrasive wear, etc., or fretting fatigue, which includes a surface or bulk fatigue. The purpose of the present work are to investigate the variation of the materials with increasing number of cycles and sliding velocity under abrasive wear test and to examine the wear mechanism at each test condition

Finite element analysis of optimized H shape spring in a nuclear fuel spacer grid by using contact definition

International Nuclear Information System (INIS)

Kim, Jae-Yong; Yoon, Kyung-Ho

2007-01-01

The primary role of the grid springs in spacer grid is to hold the fuel rods in an appropriate position using friction force and to prevent the fuel rods dropping during reactor operation. The spring force decreases as the fuel burn-up increases since the spring stiffness is degraded due to the high temperature and the irradiation effect in the reactor core. So this phenomenon has to be considered when the initial spring force of grid spring is designed. To check whether the spring have suitable spring force, the characterization test of spring is conducted. In this paper, finite element analysis using contact definition is established for prediction the spring stiffness without test. The test and analysis results are compared to check the availability of finite element model for investing the spring characteristics in assembly condition. (author)

Measurement of grid spacer's enhanced droplet cooling under reflood condition in a PWR by LDA

International Nuclear Information System (INIS)

Lee, S.L.; Sheen, H.J.; Cho, S.K.; Issapour, I.; Hua, S.Q.

1984-01-01

Reported is an experiment designed for the measurements of grid spacer's enhanced droplet cooling under reflood condition at elevated temperatures in a steam environment. The flow channel consists of a simulated 1.60m-long pressurized water reactor (PWR) fuel rod bundle of 2 x 2 electrically heated rods. Embedded thermocouples are used to measure the rod cladding temperature at various axial levels and an unshielded Chromel-Alumel thermocouple sheathed by a small Inconel tube is traversed in the center of the subchannel to measure the temperatures of the water and steam coolant phases at various levels. The droplet dynamics across the grid spacer is directly obtained by a special laser-Doppler anemometry technique for the in situ simultaneous measurement of velocity and size of droplets through two observation windows on the test channel, one immediately before and one immediately after the grid spacer. Some results are presented and analyzed

Development of MHI PWR fuel assembly with high thermal performance

International Nuclear Information System (INIS)

Yasushi Makino; Masaya Hoshi; Masaji Mori; Hidetoshi Kido; Kazuo Ikeda

2005-01-01

Mitsubishi Heavy Industries, Ltd. (MHI) has been developing a PWR fuel assembly to meet the needs of Japanese fuel market with mainly improving its reliability such as a mechanical strength, a seismic strength and endurance. For burn-up extension of the fuel to 55 GWd/t, MHI has introduced a Zircaloy spacer grid with better neutron economics with retaining the reliability in an operating core. However, for a future power up-rating and a longer cycle operation, a higher thermal performance is required for PWR fuel assembly. To meet the needs of fuel market, MHI has developed an advanced type of Zircaloy spacer grid with a greater DNB performance while retaining the reliability of a fuel and a relatively low pressure drop. For the greater DNB performance, MHI optimized geometrical shape of mixing vane to promote a fluid mixing performance. In this report, higher DNB performance provided by the advanced Zircaloy spacer grid is presented. The results of 3D simulation for the flow behavior in 5 x 5 partial assembly, a mixing test and a water DNB test were compared between the current and the advanced spacer grids. Consequently, it was confirmed that a crossover vane enhanced a fluid mixing and the advanced spacer grid could significantly improve DNB performance compared with the current design of spacer grids. (authors)

Procedure for vibration test of the fuel rod supported by spacer grids

International Nuclear Information System (INIS)

Choi, Myoung Hwan; Kang, Heung Seok; Yoon, Kyung Ho; Kim, Hyung Kyu; Song, Kee Nam

2002-07-01

One of the methods that are used to compare and verify the supporting performance of the spacer grids developed is the vibration characteristic test. In this report there are two aims. One is of the understand of the experimental method and procedure performing the modal testing using I-DEAS TDAS module. The other is the investigation of the vibration behaviors of a dummy fuel rod supported by 8 optimized H type spacer grids. This report describes the method and procedure of modal testing to obtain the vibration characteristics such as amplitudes, natural frequencies and mode shapes of the fuel rod using a shaker, a non-contact gap sensor and an accelerometer. This report provides a test procedure in detail so that anyone can be easily understood and use the I-DEAS TDAS program. The I-DEAS TDAS program related to the modal testing has several tasks including the Modal analysis, Signal Processing et al.. This report includes model preparation to prepare the geometrical model, Signal Processing (Sine/Standard measurement) to acquire the signal, Modal analysis to obtain the frequencies and mode shapes, Correlation to analyze the relation between the test and FE analysis and Post Processing tasks. In addition, this report contains the actual test and analysis data of a dummy fuel rod in length 3847mm supported by 8 optimized H type spacer grids

Finite element analysis of the contact between fuel rod and spacer grid

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyung Kyu; Kim, Young Koon; Kang, Heung Seok; Yoon, Kyung Ho; Song, Kee Nam [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-01-01

For the research on the fretting failure problem of nuclear fuel, the contact length and normal stress field are evaluated for the contact between fuel rod and spacer grid by using the Finite Element Method (FEM). An assumption of semi-infiniteness is necessary for applying the Contact Mechanics which is based on the classical theory of elasticity to the present problem. For the contact problem of fuel fretting, the contact mechanical solutions could be utilized well with sufficient accuracy if the contact bodies (i.e., the cladding tube and the spacer grid) can be assumed as semi-infinite bodies. To this end, the contact length evaluated by FEM is discussed together with the relevant research which concerned the effect of dimension for the validity of the assumption of semi-infiniteness. Normal stress profile on the contact is also studied through comparing the theoretical and the FE results. For the analysis of contact problem by FEM, ANSYS code (Version 5.3) is utilized and the geometry is chosen to be the Hertzian (cylinder-to-cylinder), the strip-to-cylinder and the fuel rod/spacer grid contact (strip-to-tube). Present research will be utilized for accessing the fuel fretting problem by FEM together with the theoretical (i.e., contact mechanical) analysis which has been published as KAERI/TR-1113/98. (author). 15 refs., 44 figs., 4 tabs.

Experimental and numerical investigation of water flow through spacer grids of nuclear fuel elements using the Open FOAM code

International Nuclear Information System (INIS)

Vidal, Guilherme A.M.; Vieira, Tiago A.S.; Castro, Higor F.P.

2017-01-01

With the advancement and development of computational tools, the studies of thermofluidodynamic behavior in nuclear fuel elements have been developed in recent years. Of the devices present in these elements, the spacing grids received more attention. They have kept the fuel rods equally spaced and have fins that aim to improve the heat transfer process between the water and the fuel element. Therefore, the grids present an important structural and thermal function. This work was carried out with the purpose of verifying and validating simulations of spacer grids using OpenFOAM (2017) software of Computational Fluid Dynamics (CFD). The simulations were validated using results obtained through the commercial CFD program, Ansys CFX, and experiments available in the literature and obtained in test sections assembled on the Water-Air Circuit (CCA) of the CDTN thermo-hydraulic laboratory

Experimental investigations of turbulent flows in rod bundles with and without spacer grids

International Nuclear Information System (INIS)

Trippe, G.

1979-07-01

In the thermofluiddynamic design of liquid metal cooled reactor fuel elements the lack of experimentally confirmed knowledge of the three-dimensional flow events in rod bundles provided with spacer grids has appeared as a significant problem. To close this gap of knowledge, detailed measurements of the local velocities were made on a 19-rod bundle model. The Pitot method of differential pressure measurements was used as the measuring system. In the first part of the work the fully developed flow regime not influenced by spacers was investigated. A simple relation was derived for distributing the mass flow among the subchannels of a rod bundle; it is but slightly dependent on the Reynolds number. This relation allows a quick, coarse calculation of the distribution of the undisturbed, fully developed mass flow in bundles with similar geometries. By evaluation of further experiments known from the literature, empirical relationships were found for the local velocity distribution within the subchannels of such bundles. In the second part the effect of grid shaped spacers was investigated. The three-dimensional flow events caused by the spacers were completely recorded and interpreted physically. The deeper understanding of these flow processes can now serve to improve the model concept used in the present design computer programs. Single results of the investigations which take primary importance are the quantitative relations existing between the changes of mass flow in the bundle boundary zone, caused by a spacer, and the geometry of this spacer. The transferability to other bundle geometries was discussed and delimited. Moreover, it was shown that the mass flow in the bundle boundary zone can be successively reduced by spacers placed one behind the other in the bundle. A noticeable dependence of flow events on the Reynolds number was not found for the range relevant in practical application (30.000 [de

RELAP5 model to simulate the thermal-hydraulic effects of grid spacers and cladding rupture during reflood

Energy Technology Data Exchange (ETDEWEB)

Nithianandan, C.K.; Klingenfus, J.A.; Reilly, S.S. [B& W Nuclear Technologies, Lynchburg, VA (United States)

1995-09-01

Droplet breakup at spacer grids and a cladding swelled and ruptured locations plays an important role in the cooling of nuclear fuel rods during the reflooding period of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). During the reflood phase, a spacer grid affects the thermal-hydraulic system behavior through increased turbulence, droplet breakup due to impact on grid straps, grid rewetting, and liquid holdup due to grid form losses. Recently, models to simulate spacer grid effects and blockage and rupture effects on system thermal hydraulics were added to the B&W Nuclear Technologies (BWNT) version of the RELAP5/MOD2 computer code. Several FLECHT-SEASET forced reflood tests, CCTF Tests C1-19 and C2-6, SCTF Test S3-15, and G2 Test 561 were simulated using RELAP5/MOD2-B&W to verify the applicability of the model at the cladding swelled and rupture locations. The results demonstrate the importance of modeling the thermal-hydraulic effects due to grids, and clad swelling and rupture to correctly predict the clad temperature response during the reflood phase of large break LOCA. The RELAP5 models and the test results are described in this paper.

Boiling transition phenomenon in BWR fuel assemblies effect of fuel spacer shape on critical power

International Nuclear Information System (INIS)

Yamamoto, Yasushi; Morooka, Shin-ichi; Mitsutake, Toru; Yokobori, Seiichi; Kimura, Jiro.

1996-01-01

A thorough understanding of the thermal-hydraulic phenomena near fuel spacer is necessary for the accurate prediction of the critical power of BWR fuel assemblies, and is thus essential for effective developments of a new BWR fuel assembly. The main purpose of this study is to develop an accurate method for predicting the effect of spacer shapes on critical power. Tests have been conducted under actual BWR operating conditions, using an annulus flow channel consisting of a heated rod and circular-tube channel, and BWR simulated 4x4 rod bundles with heater rods unheated just upsteam of spacer. The effect of spacer shapes on critical power was predicted analytically based on the droplet deposition rate estimation. The droplet deposition rate for different spacer shapes was calculated using a single-phase flow model. The prediction results were compared with the test results for the annulus flow channel using ring-type spacers. Analytical results of critical power agreed with measured critical power from point of the effects of changes in the rod-spacer clearance and the spacer thickness on critical power. (author)

Eddy current monitoring of spacers in coolant channel assemblies of nuclear reactor

International Nuclear Information System (INIS)

Bhole, V.M.; Rastogi, P.K.; Kulkarni, P.G.; Vijayaraghavan, R.

1993-01-01

An eddy current testing method has been standardised for monitoring spacer springs which are used in coolant channel assemblies of pressurised heavy water nuclear reactors (PHWRs). The standard bobbin coil probe used for monitoring the spacer spring detects only the location but does not monitor the tilt orientation and tilt angle of a tilted spacer spring. The knowledge of location along with the tilt orientation of the spacer spring greatly improves the performance of repositioning methods. A modified probe with angular windings has been developed in laboratory tests for monitoring the location as well as the tilt orientation of the spacer springs. Experimental results are presented showing excellent performance of the modified probe in monitoring the exact location as well as tilt orientation of a spacer spring. The modified probe has also been used successfully in the field during repositioning of spacer springs in PHWRs before commissioning. (Author)

Fuel assembly guide tube

International Nuclear Information System (INIS)

Jabsen, F.S.

1979-01-01

This invention is directed toward a nuclear fuel assembly guide tube arrangement which restrains spacer grid movement due to coolant flow and which offers secondary means for supporting a fuel assembly during handling and transfer operations

An insulating grid spacer for large-area MICROMEGAS chambers

International Nuclear Information System (INIS)

Bernard, D.; Delagrange, H.; D'Enterria, D.G.; Guay, M.L.M. Le; Martinez, G.; Mora, M.J.; Pichot, P.; Roy, D.; Schutz, Y.; Gandi, A.; Oliveira, R. de

2002-01-01

We present a novel design for large-area gaseous detectors based on the MICROMEGAS technology. This technology incorporates an insulating grid, sandwiched between the micro-mesh and the anode-pad plane, which provides a uniform 200 μm amplification gap. The uniformity of the amplification gap thickness has been verified. The gain performances of the detector are presented and compared to the values obtained with detectors using cylindrical micro spacers. The new design presents several technical and financial advantages

Spacers for use in nuclear fuel assembly

International Nuclear Information System (INIS)

Shiohata, Hironori; Nakamura, Shozo; Hasegawa, Kunio; Higuchi, Shigeo; Nagashima, Hideaki; Kawada, Yoshishige.

1987-01-01

Purpose: To prevent liquid film breakage at the surface of a fuel rod due to swirlings of steam flow generated at the upstream of a contact portion between the fuel rod and a spacer leaf spring, that is, below the contact portion. Constitution: Steam-hot water 2-phase streams flowing from the lower to the upper portions of a fuel assembly is hindered by leaf springs, thereby forming swirlings in the steam flow at the upstream of a contact portion between the fuel rod and the leaf springs, that is, below the contact portion. The horseshoe-like swirlings shed the liquid films at the surface of the fuel rod to remarkably decrease the heat cooling performance, by which the surface temperature of a fuel can is temporarily increased thereby possibly causing failures due to so-called burnout in view of the above, steps are formed to the spacer leaf spring for use in the fuel assembly, to reduce the pressure difference between the leaf spring and the fuel rod at the upstream of the springs relative to the 2-phase coolant stream. In this way, formation of the swirlings is moderated to prevent the liquid film breakage and improve the critical heat power. (Kamimura, M.)

An insulating grid spacer for large-area MICROMEGAS chambers

CERN Document Server

Bernard, D; D'Enterria, D G; Le Guay, M; Martínez, G; Mora, M J; Pichot, P; Roy, D; Schutz, Y; Gandi, A; De Oliveira, R

2002-01-01

We present an original design for large area gaseous detectors based on the MICROMEGAS technology. This technology incorporates an insulating grid, sandwiched between the micro-mesh and the anode-pad plane, which provides an uniform 200 $\\mu$m amplification gap. The uniformity of the amplification gap thickness has been verified under several experimental conditions. The gain performances of the detector are presented and compared to the values obtained with detectors using cylindrical micro spacers. The new design presents several technical and financial advantages.

Nuclear fuel assembly

International Nuclear Information System (INIS)

Anthony, A.J.

1980-01-01

A bimetallic spacer means is cooperatively associated with a nuclear fuel assembly and operative to resist the occurrence of in-reactor bowing of the nuclear fuel assembly. The bimetallic spacer means in one embodiment of the invention includes a space grid formed, at least principally, of zircaloy to the external surface of which are attached a plurality of stainless steel strips. In another embodiment the strips are attached to fuel pins. In each of the embodiments, the stainless steel strips during power production expand outwardly to a greater extent than do the members to which the stainless steel strips are attached, thereby forming stiff springs which abut against like bimetallic spacer means with which the other nuclear fuel assemblies are provided in a given nuclear reactor core to thus prevent the occurrence of in-reactor bowing of the nuclear fuel assemblies. (author)

BWR fuel assembly having fuel rod spacers axially positioned by exterior springs

International Nuclear Information System (INIS)

Taleyarkhan, R.P.

1988-01-01

In a fuel assembly having spaced fuel rods, an outer hollow tubular flow channel surrounding the fuel rods so as to direct flow of coolant/moderator fluid there-along, and at least one spacer being disposed along the channel and about the fuel rods so as to maintain them in side-by-side spaced relationship, an arrangement for disposing the spacer in a desired axial position along the fuel rods is described comprising: yieldably resilient springs disposed between an interior side of the outer channel and an exterior side of the spacer. The springs have an inherent spring bias directed away from the exterior sides of the spacers and toward the interior side of the channel such that by contact with the channel and spacer the springs assume states in which they are deflected away from the channel interior side so as to exert sufficient compressive contacting force thereon to maintain the spacer substantially stationary in the desired axial position along the fuel rods

Plan for Structural Analysis of Fuel Assembly for Seismic and Loss of Coolant Accident Loading Considering End-Of-Life Condition for APR1400 NRC Design Certification

Energy Technology Data Exchange (ETDEWEB)

Kim, Dong-Hak [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The evaluation of fuel assembly structural response to externally applied forces by earthquakes and postulated pipe breaks in the reactor coolant system is described in standard review plan (SRP) 4.2, appendix A. SRP 4.2, appendix A, section III, states, 'While P(crit) [the crushing load] will increase with irradiation, ductility will be reduced. The extra margin in P(crit) for irradiated spacer grids is thus assumed to offset the unknown deformation behavior of irradiated spacer grids beyond P(crit).' The assumption in the SRP concerning irradiated grids may suggest that only the beginning-of-life (BOL) condition for spacer grid strength needs to be evaluated for fuel assembly integrity under externally applied forces. However, U.S. NRC issued the NRC. To consider the EOL conditions for the structural analysis of the fuel assembly under a seismic and LOCA loading, the simulated fuel assembly for EOL conditions should be considered by determining the gap between the spacer grid and fuel rod. Using the simulated fuel assembly, spacer grid test and fuel assembly mechanical test should be conducted to determine the simplified model of fuel assembly which is used for the structural analysis. The structural analysis will be conducted using the fuel assembly model for EOL condition. The flow damping value will be also used for the structural analysis to reduce the impact force.

System for measuring spacer pin pitch in a nuclear fuel assembly

International Nuclear Information System (INIS)

Isono, Kenji; Tateishi, Yoshinori; Mano, Tadashi.

1975-01-01

Object: To reduce the period for discriminating whether or not spacer pin pitch is satisfactory by simultaneously inserting a number of reference rods into a nuclear fuel assembly spacer ring element of a reactor and arranging them such that they can be simultaneously withdrawn to simplify the withdrawing operation. Structure: A spacer provided with a ring element which clamps a nuclear fuel element is supported on a spacer support with a rod secured to the support as a guide and is secured to the support by securing means. A vertically movable structure with a reference rod provided upright and thru-holes formed in two support plates provided in the same row as the spacer ring element is operated by a fluid pressure mechanism to simultaneously insert the reference rod into the spacer ring element. The reference rod is mounted in support plates via ball bearings such that it is slightly movable in the horizontal direction, and it is aligned with respect to the core of the ring element. The intercore distance of the reference rod is measured with the reference rod inserted in the ring element, thereby measuring the space pin pitch. From the results of measurement, discrimination as to whether the spacer is satisfactory or not is made. (Kamimura, M.)

CFD simulation and validation of turbulent mixing in a rod bundle with vaned spacer grids based on LDV test

International Nuclear Information System (INIS)

Chen Xi; Li Songwei; Li Zhongchun; Du Sijia; Zhang Yu; Peng Huanhuan

2017-01-01

Spacer grids with mixing vanes are generally used in fuel assemblies of Pressurized Water Reactor (PWR), because that mixing vanes could enhance the lateral turbulent mixing in subchannels. Thus, heat exchangements are more efficient, and the value of departure from nucleate boiling (DNB) is greatly increased. Actually turbulent mixing is composed of two kinds of flows: swirling flow inside the subchannel and cross flow between subchannels. Swirling flow could induce mixing between hot water near the rod and cold water in the center of the subchannel, and may accelerate deviation of the bubbles from the rod surface. Besides, crossing flow help to mixing water between hot subchannels and cold subchannels, which impact relatively large flow area. As a result, how to accurately capture and how to predict the complicated mixing phenomenon are of great concernments. Recently many experimental studies has been conducted to provide detailed turbulent mixing in rod bundle, among which Laser Doppler Velocimetry method is widely used. With great development of Computational Fluid Dynamics, CFD has been validated as an analysis method for nuclear engineering, especially for single phase calculation. This paper presents the CFD simulation and validation of the turbulent mixing induced by spacer grid with mixing vanes in rod bundles. Experiment data used for validation came from 5 x 5 rod bundle test with LDV technology, which is organized by Science and Technology on Reactor System Design Technology Laboratory. A 5 x 5 rod bundle with two spacer grids were used. Each rod has dimension of 9.5 mm in outer diameter and distance between rods is 12.6 mm. Two axial bulk velocities were conducted at 3.0 m/s for high Reynolds number and 1.0 m/s for low Reynolds number. Working pressure was 1.0 bar, and temperature was about 25degC. Two different distances from the downstream of the mixing spacer grid and one from upstream were acquired. Mean axial velocities and turbulent intensities

Experimental studies of resistance fretting-wear of fuel rods for VVER-1000 and TVS-KVADRAT fuel assemblies

International Nuclear Information System (INIS)

Makarov, V.; Afanasiev, A.; Egorov, Yu.; Matvienko, I.

2015-01-01

The paper covers the results of the studies performed to justify the wear resistance of fuel rods in contact with the spacer grids of TVS VVER-1000 fuel assembly and TVS-KVADRAT square fuel assembly of Russian design for PWR-900 reactor. The presented results of three testing stages comprise: Testing of mockup fuel rods of VVER TVS fuel assembly for fretting wear under the conditions of the water chemistry of VVER reactor; Testing models of different design embodiments of the fuel rods for VVER TVS fuel assembly for fretting wear in still cold water; Testing mockup fuel rods of TVS-KVADRAT square fuel assembly for PWR reactor for frettingwear under the conditions of PWR water chemistry. The effect of structural and operational factors was determined (amplitudes, fuel rod vibration frequencies, values of cladding-to-spacer grid cell gap for the depth of fuel rod cladding wear etc.), an assessment was made of the threshold values of fuel rod vibration parameters, which, if not exceeded, provide the absence of the fuel rod cladding fretting wear in the fuel rod-to spacer grid contact area. Key words: fretting wear, fuel rod, spacer grid, VVER, PWR (author)

Computational model for turbulent flow around a grid spacer with mixing vane

International Nuclear Information System (INIS)

Tsutomu Ikeno; Takeo Kajishima

2005-01-01

Turbulent mixing coefficient and pressure drop are important factors in subchannel analysis to predict onset of DNB. However, universal correlations are difficult since these factors are significantly affected by the geometry of subchannel and a grid spacer with mixing vane. Therefore, we propose a computational model to estimate these factors. Computational model: To represent the effect of geometry of grid spacer in computational model, we applied a large eddy simulation (LES) technique in couple with an improved immersed-boundary method. In our previous work (Ikeno, et al., NURETH-10), detailed properties of turbulence in subchannel were successfully investigated by developing the immersed boundary method in LES. In this study, additional improvements are given: new one-equation dynamic sub-grid scale (SGS) model is introduced to account for the complex geometry without any artificial modification; the higher order accuracy is maintained by consistent treatment for boundary conditions for velocity and pressure. NUMERICAL TEST AND DISCUSSION: Turbulent mixing coefficient and pressure drop are affected strongly by the arrangement and inclination of mixing vane. Therefore, computations are carried out for each of convolute and periodic arrangements, and for each of 30 degree and 20 degree inclinations. The difference in turbulent mixing coefficient due to these factors is reasonably predicted by our method. (An example of this numerical test is shown in Fig. 1.) Turbulent flow of the problem includes unsteady separation behind the mixing vane and vortex shedding in downstream. Anisotropic distribution of turbulent stress is also appeared in rod gap. Therefore, our computational model has advantage for assessing the influence of arrangement and inclination of mixing vane. By coarser computational mesh, one can screen several candidates for spacer design. Then, by finer mesh, more quantitative analysis is possible. By such a scheme, we believe this method is useful

Steady-state, local temperature fields with turbulent liquid sodium flow in nominal and disturbed bundle geometries with spacer grids

International Nuclear Information System (INIS)

Moeller, R.; Tschoeke, H.

1980-01-01

The operating reliability of nuclear reactors calls for a reliable strength analysis of the highly loaded core elements, one of its prerequisites being the reliable determination of the three-dimensional velocity and temperature fields. To verify thermohydraulics computer programs, extensive local temperature measurements in the rod claddings of the critical bundle zone were performed on a heated 19-rod bundle model with sodium flow and provided with spacer grids (P/D = 1.30; W/D = 1.19). The essential results are: - Outside the spacer grids, the azimuthal temperature variations of the side and corner rods are approximately 10-fold those of rods in the central bundle zone. - The spacer grids investigated give rise to great local temperature peaks and correspondingly great temperature gradients in the axial and azimuthal directions immediately around the support points. - Continuous reduction of a subchannel by rod bowing results in substantial rises of temperature which, however, are limited to adjacent cladding tubes. (orig.)

Steady-state, local temperature fields with turbulent sodium flow in nominal and disturbed bundle geometries with spacer grids

International Nuclear Information System (INIS)

Moeller, R.; Tschoeke, H.; Kolodziej, M.

1980-12-01

The operating reliability of nuclear reactors calls for a reliable strength analysis of the highly loaded core elements, one of its prerequisites being the reliable determination of the three-dimensional velocity and temperature fields. To verify thermohydraulics computer programs, extensive local temperature measurements in the rod claddings of the critical bundle zone were performed on a heated 19-rod bundle model with sodium flow and provided with spacer grids (P/D = 1.30; W/D = 1.19). These are the essential results obtained: Outside the spacer grids the azimuthal temperature variations of the side and corner rods are greater by approximately the factor 10 in the bundle geometry under consideration as compared to rods in the central bundle zone. The spacer grids investigated give rise to great local temperature peaks and correspondingly great temperature gradients in the axial and azimuthal directions immediately around the support points. Continuous reduction of a subchannel by rod bowing results in substantial rises of temperature which, however, are limited to the adjacent cladding tube zones. (orig.) [de

An Investigation on Irradiation-induced Grid Width Growth in Advanced Fuels

International Nuclear Information System (INIS)

Jang, Young Ki; Jeon, Kyeong Lak; Kim, Yong Hwan; Kim, Jae Ik; Hwang, Sun Tack; Kim, Man Su; Lee, Tae Hyoung; Yoo, Myeong Jong; Yoon, Yong Bae; Kim, Tae Wan

2011-01-01

The spacer grids for fuel assembly are fabricated from preformed Zircaloy or Inconel strips interlocked in an egg crate fashion and welded or brazed together. The spacer grid is the important component to maintain the fuel rod array by providing positive lateral restraint to the fuel rods but only frictional restraint to axial fuel rod motion. To improve economy and safety aspects, advanced nuclear fuels of PLUS7, 16ACE7 and 17ACE7 were developed. The former is for Optimized Power Reactor of 1000 MWe (OPR1000) and Advanced Power Reactor of 1400 MWe (APR1400) and the latter two are for 16x16 and 17x17 Westinghouse type reactors, respectively. The material for top and bottom spacer grids on these advanced fuels are Inconel and the mid grids are Zirlo patented by Westinghouse. For neutron economy, the fuel assemblies are arranged very closely and the gaps between assemblies are kept to around 1 mm based on the worst case. The Zirconium-based alloys grow during irradiation in reactor. The large growth may cause some difficulties in loading and unloading fuel assemblies during refueling outage in reactor. The severe growth may cause some problems that fuel assemblies may be stuck within the core shroud and a modification of loading pattern is required. In addition, the grid growth with grid spring relaxation may cause different rod vibration behavior and results in the different wear mechanism. The grid width growth on the advanced fuels were predicted by using the growth models before the irradiation in reactor and were examined using lead test assemblies (LTAs) after each cycle in Ulchin unit 3 and Kori units 2 and 3, respectively. To reconfirm irradiation performance results using LTAs, the additional examinations are being performed through the surveillance programs on the commercially supplied fuels in Yonggwang unit 5 and Kori units 2 and 4. It is investigated on this study whether the grid widths on the advanced fuels meet their criteria and the predicted models

FAMREC, PWR Lateral Mechanical Fuel Rod Assembly Response

International Nuclear Information System (INIS)

Guenzler, R.C.

1995-01-01

1 - Description of program or function: The Fuel Assembly Mechanical Response Code (FAMREC) calculates the lateral mechanical response of a row of fuel assemblies while allowing for two types of nonlinearities. The first type is a geometric nonlinearity in the form of gaps between individual assemblies and between peripheral assemblies and a boundary wall. Impacting is monitored across the gaps. The second nonlinearity is the permanent deformation of the fuel assembly spacer grid to compressive loading. 2 - Method of solution: The response is calculated in the modal plane. The coupled differential equations are solved in closed form using Laplace transformations. The discrete displacements and velocities are then calculated and the gaps in the system monitored at each axial elevation for impacting. These impact forces are then applied statistically at a given time-step, and equilibrium is found using a Gaussian elimination technique. Three impact force calculation methods are available: 1- a linear impact force and crushing load audit calculation, 2- a more detailed linear impact force and crushing load calculation, and 3- a non-linear grid calculation which allows for plastic deformation of the fuel assembly spacer grids. 3 - Restrictions on the complexity of the problem: Maxima of: 3601 time-steps and forces; 80 modes; 30 applied forces; 15 fuel assemblies; and 5 impact grids per assembly

The effect of the fuel rod friction force to the fuel assembly lateral mechanical characteristics

International Nuclear Information System (INIS)

Ha, Dong Geun; Jeon, Sang Youn; Suh, Jung Min

2012-01-01

The Fuel Assembly (FA) for light water reactor consists of hundreds of fuel rods, guide tubes, spacer grids, top/bottom nozzles. The guide tubes transmit vertical loads between the top and bottom nozzles, position the fuel rod support grids vertically, react the loads from the fuel rods that are applied to the grids, and provide some of the lateral load capability for the overall fuel assembly. The guide tubes are the structural members of the skeleton assembly. And the spacer grids maintain the fuel rod array by providing positive lateral restraint to the fuel rod but only frictional restraint in the axial direction. Figure 1 shows the outline of skeleton, FA and the location of guide tubes in the view of cross section. 17x17 FA has 24 guide tubes and one instrumentation tube. When the FA is in reactor, the lateral stiffness is one of very important factors from the view point of in reactor integrity of fuel assembly such as guarantee of the cool able geometry, the control rod insertion etc. The lateral stiffness of FA is mainly determined by skeleton lateral stiffness. And the fuel rods loaded in the spacer grids reinforce the FA lateral stiffness. Generally, fuel rods and spacer grids create the nonlinear friction force between fuel rod tube and grid spring/dimple against external lateral force of FA. Thus, it is necessary to study the contribution of the fuel rods friction force to the FA lateral stiffness. So, this paper is to show how much amount of the fuel rod grid interaction contributes to the FA lateral stiffness based on the test results

Study of the influence of temperature and time on the electroplating nickel layer in Inconel 718 strips used in spacer grid of Pressurized Water Cooled nuclear reactors (PWR)

Energy Technology Data Exchange (ETDEWEB)

Rezende, Renato; Abati, Amanda; Verne, Júlio; Panossian, Zehbour, E-mail: amanda.abati@marinha.mil.br, E-mail: jvernegropp@gmail.com, E-mail: renato.rezende@marinha.mil.br, E-mail: zep@ipt.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil). Laboratório de Desenvolvimento e Instrumentação de Combustível Nuclear; Instituto de Pesquisas Tecnológicas (IPT), São Paulo, SP (Brazil)

2017-07-01

The Inconel 718 (UNS N07718: Ni-{sup 19}Cr-{sup 18}Fe-{sup 5}Nb-3 Mo) is a precipitation hardenable nickel alloy that has good corrosion resistance and high mechanical strength. These strips are used for assembling the spacer grid of fuel element of pressurized water cooled nuclear reactors (PWR). The spacer grid is a structural component of fundamental importance in fuel elements of PWR reactors, maintaining the position and necessary spacing of the fuel rods within the arrangement of the fuel element. The spacer grid is formed by joining the points of intersection of the strips, by a joint process called brazing. For this process, these strips are stamped and plated with a thin layer of nickel by means of electroplating in order to protect against oxidation and allow a better flowability and wettability of the addition metal in the strips during brazing. Oxidation at the surface of the base material harms wettability and inhibits spreading of the liquid addition metal on the substrate surface during the brazing process. The use of coatings such as nickel plating is used to ensure such conditions. The results showed that there is a process of diffusion de some chemical elements such as chromium, iron, titanium and aluminum from the substrate to the nickel layer and nickel from the layer to the substrate. These chemical elements are responsible for the oxidation at the surface of the strip. (author)

Characterization test of the optimized H type grid spring under the in-grid boundary condition

International Nuclear Information System (INIS)

Yoon, Kyung Ho; Kim, Hyung Tae; Lee, Kang Hee; Kang, Heung Seok

2004-01-01

This paper is dealing with the simplified method for the characterization test. In reality, the spring should be tested with the test specimen of spacer grid assembly itself (assembly-based test). But the available space for loading is very small due to the dimples and the spring of the adjacent straps, much care should be necessary in conducting the assembly-based test. Test results are compared with the previous test results for checking the usability of the assembly-based test

Experimental observation of the droplet size change across a wet grid spacer in a 6 × 6 rod bundle

International Nuclear Information System (INIS)

Cho, Hyoung Kyu; Choi, Ki Yong; Cho, Seok; Song, Chul-Hwa

2011-01-01

Highlights: ► In this study, an experiment on the droplet behavior inside a heated rod bundle has been performed. ► The experiment was focused on the change of droplet size induced by a spacer grid in a rod bundle. ► The major measuring parameters of the experiment were the droplet size and velocity. ► This test provided the data on the change of the droplet size after collision with a wet grid spacer. - Abstract: During the reflood phase of a postulated loss of coolant accident in a nuclear reactor, entrainment of liquid droplets can occur at a quench front of reflooding water. It is widely recognized that the behavior of the entrained droplets crucially affects the reflood heat transfer phenomena by decreasing the superheated steam temperature and interacting with a rod bundle and spacer grids. For this reason, various experimental and numerical studies have been performed to examine droplet behavior such as the droplet size, velocity and droplet fraction inside a rod array. In this study, an experiment on the droplet behavior inside a heated rod bundle has been performed. The experiment was focused on the change of droplet size induced by a spacer grid in a rod bundle geometry, which results in the change of the interfacial heat transfer between droplets and superheated steam. A 6 × 6 rod bundle test facility in Korea Atomic Energy Research Institute was used for the experiment. Steam was supplied by an external boiler into the bottom of the test channel, and a droplet injection nozzle was equipped instead of simulating a quench front of reflooding water. The major measuring parameters of the experiment were the droplet size and velocity, which were measured by a high-speed camera and a digital image processing technique. A series of experiments were conducted with various flow conditions of a steam injection velocity, heater temperature, droplet size, and droplet flow rate. The experiments provided the data on the change of the Sauter mean diameter of

A partial grid for a nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Demario, E.E.

1985-01-01

The invention relates to a nuclear-reactor fuel assembly including fuel-rod supporting transverse grids. The fuel assembly includes at least one additional transverse grid which is disposed between two fuel-rod supporting grids and consists of at least one partial grid structure extending across only a portion of the fuel assembly and having fuel rods and control-rod guide thimbles of only said portion extending therethrough. The partial grid structure includes means for providing lateral support of the fuel rods and/or means for laterally deflecting coolant flow, and it is formed of inter-leaved inner straps and border straps, the interleaved inner straps preferably being of substantially smaller height than the border straps to reduce the amount of material capable of parasitically absorbing neutrons. The additional transverse grid may comprise several partial grid structures associated with different groups of fuel rods of the fuel assembly

The mechanical measurement for reactor grid spacers by the laser speckle photography methd

International Nuclear Information System (INIS)

Yang Guoping; Chen Shitong.

1987-01-01

This paper describes the process of displacement, which was measured by laser speckle photography method, of reactor grid spacers during the vertical loading time. (from starting loading up to buckling). The internal forces and critical stresses were calculated by measured displacement. The effect of initial processed eccentric was considered by adding an internal force. The value of theoretical buckling load was in good agreement with the experimental result

Fuel spacer

International Nuclear Information System (INIS)

Nishida, Koji; Yokomizo, Osamu; Kanazawa, Toru; Kashiwai, Shin-ichi; Orii, Akihito.

1992-01-01

The present invention concerns a fuel spacer for a fuel assembly of a BWR type reactor and a PTR type reactor. Springs each having a vane are disposed on the side surface of a circular cell which supports a fuel rods. A vortex streams having a vertical component are formed by the vanes in the flowing direction of a flowing channel between adjacent cylindrical cells. Liquid droplets carried by streams are deposited on liquid membrane streams flowing along the fuel rod at the downstream of the spacer by the vortex streams. In view of the above, the liquid droplets can be deposited to the fuel rod without increasing the amount of metal of the spacer. Accordingly, the thermal margin of the fuel assembly can be improved without losing neutron economy. (I.N.)

Analysis of US patents on spacer grids

International Nuclear Information System (INIS)

Kim, Hyung Kyu; Song, Kee Nam; Yoon, Kyung Ho; Kang, Hong Seok; Kim, Hyung Kyu; Jeon, Tae Hyun; Oh, Dong Seok; In, Wang Ki; Bang, Jae Keun; Oh, Seung Eun; Seo, Jeong Min; Lee, Jin Seok; Park, Seong Keun

1997-06-01

The total of 137 US patents on spacer grids patented from 1968 through 1993 are analyzed and summarized. Database is constituted with designing the appropriate fields by which each patent can be identified. The fields consist of patent number, inventor, assignee, date of patent, title and major foci of the patent. The major foci are again classified by detailed subjects such as the fretting failure and fuel rod support-related, the strength-related, the fabrication-related as for mechanical subjects, while the cooling performance-related and the pressure drop-related as for thermal-hydraulic one. The 92% of the patents analyzed were issued form nuclear companies of USA, France and Germany. Among the patents dealing with mechanical subjects, the fretting failure and fuel rod support-related is more than the pressure drop-related among the patents of thermal-hydraulic subjects. The number of patents issued from Japan ranks just after Germany i.e., the 4th country. It is thought that much concern as well as investment should be increased in this field, the patent of nuclear components. (author). 2 tabs., 5 figs

Analysis of US patents on spacer grids

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyung Kyu; Song, Kee Nam; Yoon, Kyung Ho; Kang, Hong Seok; Kim, Hyung Kyu; Jeon, Tae Hyun; Oh, Dong Seok; In, Wang Ki; Bang, Jae Keun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Oh, Seung Eun; Seo, Jeong Min; Lee, Jin Seok; Park, Seong Keun [Korea Nuclear Fuel Company, Taejon (Korea, Republic of)

1997-06-01

The total of 137 US patents on spacer grids patented from 1968 through 1993 are analyzed and summarized. Database is constituted with designing the appropriate fields by which each patent can be identified. The fields consist of patent number, inventor, assignee, date of patent, title and major foci of the patent. The major foci are again classified by detailed subjects such as the fretting failure and fuel rod support-related, the strength-related, the fabrication-related as for mechanical subjects, while the cooling performance-related and the pressure drop-related as for thermal-hydraulic one. The 92% of the patents analyzed were issued form nuclear companies of USA, France and Germany. Among the patents dealing with mechanical subjects, the fretting failure and fuel rod support-related is more than the pressure drop-related among the patents of thermal-hydraulic subjects. The number of patents issued from Japan ranks just after Germany i.e., the 4th country. It is thought that much concern as well as investment should be increased in this field, the patent of nuclear components. (author). 2 tabs., 5 figs.

Development of CFD analysis method based on droplet tracking model for BWR fuel assemblies

International Nuclear Information System (INIS)

Onishi, Yoichi; Minato, Akihiko; Ichikawa, Ryoko; Mashara, Yasuhiro

2011-01-01

It is well known that the minimum critical power ratio (MCPR) of the boiling water reactor (BWR) fuel assembly depends on the spacer grid type. Recently, improvement of the critical power is being studied by using a spacer grid with mixing devices attaching various types of flow deflectors. In order to predict the critical power of the improved BWR fuel assembly, we have developed an analysis method based on the consideration of detailed thermal-hydraulic mechanism of annular mist flow regime in the subchannels for an arbitrary spacer type. The proposed method is based on a computational fluid dynamics (CFD) model with a droplet tracking model for analyzing the vapor-phase turbulent flow in which droplets are transported in the subchannels of the BWR fuel assembly. We adopted the general-purpose CFD software Advance/FrontFlow/red (AFFr) as the base code, which is a commercial software package created as a part of Japanese national project. AFFr employs a three-dimensional (3D) unstructured grid system for application to complex geometries. First, AFFr was applied to single-phase flows of gas in the present paper. The calculated results were compared with experiments using a round cellular spacer in one subchannel to investigate the influence of the choice of turbulence model. The analyses using the large eddy simulation (LES) and re-normalisation group (RNG) k-ε models were carried out. The results of both the LES and RNG k-ε models show that calculations of velocity distribution and velocity fluctuation distribution in the spacer downstream reproduce the experimental results qualitatively. However, the velocity distribution analyzed by the LES model is better than that by the RNG k-ε model. The velocity fluctuation near the fuel rod, which is important for droplet deposition to the rod, is also simulated well by the LES model. Then, to examine the effect of the spacer shape on the analytical result, the gas flow analyses with the RNG k-ε model were performed

Numerical Analysis for IFM Grid Effect on 5x5 Rods Bundle

International Nuclear Information System (INIS)

Kim, Seong Jin; Cha, Jeong Hun; Seo, Kyong Won; Kim, Tae Woo; Kwon, Hyuk; Hwang, Dae Hyun

2011-01-01

Generally, the fuel assembly consists of fuel rods, bottom and top grids, spacer grids, mixing vane, etc. The mixing vane with spacer grid is used to increase the thermal mixing between subchannels and to increase CHF(Critical Heat Flux). IFM(Intermediate Flow Mixer) grids are used to induce lateral flow between adjacent channels and are well-known as improving CHF, also. A numerical analysis using CFD code(ANSYS CFX, version 12.1) and subchannel code(MATRA-S) was conducted to investigate the influence of IFM grid on the subchannel temperature in 5x5 rods bundle with and without the IFM grid, thermohydraulically. In this study, the quantitative improvement of the mixing effect of the IFM grid is presented from the results of CFX and MATRA-S code. Moreover, capacity of predicting subchannel temperature of MATRA-S code is compared with CFX result

Fuel assembly supporting structure

International Nuclear Information System (INIS)

Aisch, F.W.; Fuchs, H.P.; Knoedler, D.; Steinke, A.; Steven, J.

1976-01-01

For use in forming the core of a pressurized-water reactor, a fuel assembly supporting structure for holding a bundle of interspaced fuel rods, is formed by interspaced end pieces having holes in which the end portions of control rod guide tubes are inserted, fuel rod spacer grids being positioned by these guide tubes between the end pieces. The end pieces are fastened to the end portions of the guide tubes, to integrate the supporting structure, and in the case of at least one of the end pieces, this is done by means which releases that end piece from the guide tubes when the end pieces receive an abnormal thrust force directed towards each other and which would otherwise place the guide tubes under a compressive stress that would cause them to buckle. The spacer grids normally hold the fuel rods interspaced by distances determined by nuclear physics, and buckling of the control rod guide tubes can distort the fuel rod spacer grids with consequent dearrangement of the fuel rod interspacing. A sudden loss of pressure in a pressurized-water reactor pressure vessel can result in the pressurized coolant in the vessel discharging from the vessel at such high velocity as to result in the abnormal thrust force on the end pieces of each fuel assembly, which could cause buckling of the control rod guide tubes when the end pieces are fixed to them in the normal rigid and unyielding manner

Sub-15 nm nano-pattern generation by spacer width control for high density precisely positioned self-assembled device nanomanufacturing

KAUST Repository

Rojas, Jhonathan Prieto; Hussain, Muhammad Mustafa

2012-01-01

We present a conventional micro-fabrication based thin film vertical sidewall (spacer) width controlled nano-gap fabrication process to create arrays of nanopatterns for high density precisely positioned self-assembled nanoelectronics device integration. We have used conventional optical lithography to create base structures and then silicon nitride (Si 3N4) based spacer formation via reactive ion etching. Control of Si3N4 thickness provides accurate control of vertical sidewall (spacer) besides the base structures. Nano-gaps are fabricated between two adjacent spacers whereas the width of the gap depends on the gap between two adjacent base structures minus width of adjacent spacers. We demonstrate the process using a 32 nm node complementary metal oxide semiconductor (CMOS) platform to show its compatibility for very large scale heterogeneous integration of top-down and bottom-up fabrication as well as conventional and selfassembled nanodevices. This process opens up clear opportunity to overcome the decade long challenge of high density integration of self-assembled devices with precise position control. © 2012 IEEE.

Sub-15 nm nano-pattern generation by spacer width control for high density precisely positioned self-assembled device nanomanufacturing

KAUST Repository

Rojas, Jhonathan Prieto

2012-08-01

We present a conventional micro-fabrication based thin film vertical sidewall (spacer) width controlled nano-gap fabrication process to create arrays of nanopatterns for high density precisely positioned self-assembled nanoelectronics device integration. We have used conventional optical lithography to create base structures and then silicon nitride (Si 3N4) based spacer formation via reactive ion etching. Control of Si3N4 thickness provides accurate control of vertical sidewall (spacer) besides the base structures. Nano-gaps are fabricated between two adjacent spacers whereas the width of the gap depends on the gap between two adjacent base structures minus width of adjacent spacers. We demonstrate the process using a 32 nm node complementary metal oxide semiconductor (CMOS) platform to show its compatibility for very large scale heterogeneous integration of top-down and bottom-up fabrication as well as conventional and selfassembled nanodevices. This process opens up clear opportunity to overcome the decade long challenge of high density integration of self-assembled devices with precise position control. © 2012 IEEE.

Preliminary results for validation of Computational Fluid Dynamics for prediction of flow through a split vane spacer grid

International Nuclear Information System (INIS)

Rashkovan, A.; Novog, D.R.

2012-01-01

This paper presents the results of the CFD simulations of turbulent flow past spacer grid with mixing vanes. This study summarizes the first stage of the ongoing numerical blind exercise organized by OECD-NEA. McMaster University along with other participants plan to submit a numerical prediction of the detailed flow field and turbulence characteristics of the flow past 5x5 rod bundle with a spacer grid equipped with two types of mixing vanes. The results will be compared with blind experimental measurements performed in Korea. Due to the fact that a number of the modeling strategies are suggested in literature for such types of flows, we have performed a series of tests to assess the mesh requirements, flow steadiness, turbulence modeling and wall treatment effects. Results of these studies are reported in the present paper. (author)

Spacer st4ructure

International Nuclear Information System (INIS)

Masetti, W.R.

1978-01-01

A spacer structure is described for maintaining a spaced relation between a plurality of generally parallel fuel rods within a housing in a nuclear reactor. The spacer structure is comprised of a grid pattern of ribs slotted to interlock with each other. The slots are arranged in such a way that when the ribs are welded to each other, the weld shrinkage is distributed uniformly in all directions to reduce or eliminate the amount of rework necessary in manufacturing the spacer structure

Investigating the thermal hydraulic performance of spacer grid with mixing vanes using STAR-CCM+ and MATRA

International Nuclear Information System (INIS)

Agbodemegbe, V. Y.

2014-07-01

Enhancement of heat transfer for flow through rod bundles is linked to the extent and sustainability of mixing in the flow geometry. Spacer grids used as support for rod bundles in nuclear reactors, when attached with ditferent designs of mixing vanes promote turbulent mixing by inducing swirl or forced lateral convection that improves mixing within or between sub-channels. The improved turbulent mixing raises the margin of the onset of critical heat flux in light water reactors (LWR) and also ensures a higher fuel cycle economy. To optimize design of mixing vanes and performance of spacer grids with mixing vanes, computational fluid dynamic simulations arc carried out on new designs and validated experimentally prior to industrial application. In the present study, computational fluid dynamic simulation using STAR-CCM+ and sub-channel analysis with MATRA were performed for flow of water through a 5 X 5 rod bundle geometry for which the rod to rod pitch to diameter ratio was 1.33 and the wall to rod pitch to diameter ratio was 0.74. The two layer k-epsilon turbulence model with an all- y + automatic wall treatment function in STAR-CCM+ were adopted for an isothermal single phase flow through the geometry with imposed cyclic periodic and non-cyclic periodic interface boundary conditions. The objective was to primarily investigate the detail flow behavior in rod bundle in the presence of spacer grids with and without attached mixing vanes. Furthermore, the present study also investigated the extent of turbulent mixing and lateral mass flux induced by the mixing vanes through comparative analysis of parametric trends for spacer grid with and without mixing vanes. New models and empirical correlations for describing the mixing vanes effects were also developed. These semi-empirical correlations improved the prediction for lateral mass flux due to turbulence, fraction of flow diverted through gaps and cross-flow resistance coefficients. Validation of simulation results

A Numerical Analysis on the Local Deformation of a Spacer Grid Structure for Nuclear Fuel Cells

International Nuclear Information System (INIS)

Jang, Myung-Geun; Na, Geum Ju; Kim, Jong-Bong; Shin, Hyunho

2016-01-01

The result of a preliminary numerical investigation on local deformation characteristics of a multi-layered spacer-grid structure with five guide tubes is reported based on implicit finite element analysis. For the numerical analysis, displacements of top and bottom cross sections of each guide tube in a single-layer model were constrained while a lateral displacement was imposed on the single layer. Unlike the impact hammer test that is generally employed to characterize the deformation characteristics of the space-grid structure, the buckling phenomenon occurs locally in this study; it takes place at the inner grids around each tube and the degree of bucking is more apparent for tubes near the lateral surface where the lateral displacement was imposed. (paper)

Vibration analysis of a dummy fuel rod continuously supported by spacer grids

International Nuclear Information System (INIS)

Choi, Myoung-Hwan; Kang, Heung-Seok; Yoon, Kyung-Ho; Song, Kee-Nam; Jung, Youn-Ho

2003-01-01

A modal testing and a finite element (FE) analysis using ABAQUS on a dummy fuel rod continuously supported by Optimized H type (OHT) and New Doublet (ND) spacer grids are performed to obtain the vibration characteristics such as natural frequencies and mode shapes and to verify the FE model used. The results from the test and the FE analysis are compared according to modal assurance criteria values. The natural frequency differences between the two methods as well as the mode comparison results for the rod with OHT SG are better than those with ND SG. That is, in the case of the ND grid model using beam-spring elements, there was a large discrepancy between the two methods. Thus, we tried to modify the FE model for ND SG considering the contact phenomena between the fuel rod and the SG. The results of the new model showed good agreement with the experiment compared with those of a beam-spring model

BWR fuel assembly with improved spacer and fuel bundle design for enhanced thermal-hydraulic performance

International Nuclear Information System (INIS)

Mildrum, C.M.; Taleyarkhan, R.P.

1987-01-01

In a fuel assembly having a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods, an outer tubular flow channel surrounding the fuel rods so as to direct flow of coolant/moderator fluid along the fuel rods, a hollow water cross extending centrally through and interconnected with the outer flow channel so as to divide the channel into separate compartments and the bundle of fuelrods into a plurality of mini-bundles thereof being disposed in the compartments, and spacers axially displaced along the fuel rods in each of the mini-bundles thereof. Each spacer is composed of inner and outer means which together define spacer cells at corner, side and interior locations of the spacer and have respective protrusions formed thereon which extend into cells so as to maintain the fuel rods received through the spacer cells in laterally spaced relationships. The improvement is described which comprises: (a) a generally uniform poison coating within at least a majority of the fuel rods; (b) a predetermined pattern of fuel enrichment with respect to the fuel rods of each mini-bundle thereof which together with the uniform poison coating within the fuel rods ensures that the packing powers of the fuel rods in the corner and side cells of the spacers are less than the peaking power of a leading one of the fuel rods in the interior cells of the spacers; and (c) each of the fuel rods being received through the cells of each spacer having a diametric size smaller than that of each of the fuel rods received through the side and interior cells of each spacer, the diametric sizes of each of the fuel rods received through the side and interior cells of each spacer being generally equal

Fabrication development of full-sized components for GCFR core assemblies

International Nuclear Information System (INIS)

Lindgren, J.R.; Flynn, P.W.; Foster, L.C.

1980-05-01

This paper presents the status of the development of full-sized components for gas-cooled fast reactor (GCFR) core assemblies. Methods for ribbing of the fuel rod cladding, fabrication of grid spacers of two different designs, drawing of assembly flow ducts, and fabrication of fission gas collection manifolds by several methods are discussed

Design of a nuclear fuel rod support grid using axiomatic design

International Nuclear Information System (INIS)

Song, Kee Nam; Yoon, Kyung Ho; Kang, Byung Soo; Park, Gyung Jin; Choi, Sung Kyoo

2002-01-01

Recently, much attention is imposed on the design of the fuel assemblies in the Pressurized Light Water Reactor (PWR). Spacer grid is one of the main structural components in a fuel assembly. It supports fuel rods, guides cooling water, and maintains a coolable geometry from the external impact loads. In this research, a new shape of the spacer grid is designed by the axiomatic approach. The Independence axiom is utilized for the design. For conceptual design, functional requirements (FRs) are defined and corresponding design parameters (DPs) are found to satisfy FRs in sequence. Overall configuration and shapes are determined in this process. Detail design is carried out based on the result of the axiomatic design. For the detail design, the system performances are evaluated by using linear and nonlinear finite element analysis. The dimensions are determined by optimization. Some commercial codes are utilized for the analysis and design

Numerical Study for Turbulent Heat Transfer in Helical Wired Sub-channel Flow Regime of Duct-less Assembly in SFR

International Nuclear Information System (INIS)

You, Byunghyun; Jeong, Yong Hoon

2014-01-01

A fuel assembly had hexagonal structure adjacent to 6 fuel assemblies, which influence to the target fuel assembly due to elimination of duct. For calculating the influence, 6 additional channels were generated between the adjacent fuel assemblies and cross flow model was applied to the channels. The adjacent fuel assemblies were analyzed and the results were used in the additional channel as boundary condition of the target fuel assembly. To design the specifications of duct-less assembly, modified or brand-new thermal-hydraulic methodology is needed which is using MATRA-LMR and CFD codes in this study. The MATRA-LMR is a sub-channel analysis code for LMR that has been developed in Korea Atomic Energy Research Institute. It is designed to analyze a fuel assembly with wire-wrap and duct structure. However, the duct-less core is not able to be analyzed by the MATRA-LMR which doesn't consider cross flow between the fuel assemblies and effect of grid spacer. The code need improvement by editing source code to eliminate effect of duct and analyze pressure drop and mixing between the sub-channels caused by grid spacer and cross flow between the fuel assemblies. To validate reformed pressure drop model and cross flow model in MATRA-LMR, CFD analysis is performed. For verifying the results of CFD, LMR subchannel experimental data is benchmarked which is done by ORNL. The verified CFD for thermalhydraulic analysis calculated pressure drop and mixing caused by grid spacer and cross flow between fuel assemblies

Measurement and CFD calculation of spacer loss coefficient for a tight-lattice fuel bundle

International Nuclear Information System (INIS)

In, Wang Kee; Shin, Chang Hwan; Kwack, Young Kyun; Lee, Chi Young

2015-01-01

Highlights: • Experiment and CFD analysis evaluated the pressure drop in a spacer grid. • The measurement and CFD errors for the spacer loss coefficient were estimated. • The spacer loss coefficient for the dual-cooled annular fuel bundle was determined. • The CFD prediction agrees with the measured spacer loss coefficient within 8%. - Abstract: An experiment and computational fluid dynamics (CFD) analysis were performed to evaluate the pressure drop in a spacer grid for a dual-cooled annular fuel (DCAF) bundle. The DCAF bundle for the Korean optimum power reactor (OPR1000) is a 12 × 12 tight-lattice rod array with a pitch-to-diameter ratio of 1.08 owing to a larger outer diameter of the annular fuel rod. An experiment was conducted to measure the pressure drop in spacer grid for the DCAF bundle. The test bundle is a full-size 12 × 12 rod bundle with 11 spacer grid. The test condition covers a Reynolds number range of 2 × 10 4 –2 × 10 5 by changing the temperature and flow rate of water. A CFD analysis was also performed to predict the pressure drop through a spacer grid using the full-size and partial bundle models. The pressure drop and loss coefficient of a spacer grid were predicted and compared with the experimental results. The CFD predictions of spacer pressure drop and loss coefficient agree with the measured values within 8%. The spacer loss coefficient for the DCAF bundle is estimated to be approximately 1.50 at a nominal operating condition of OPR1000, i.e., Re = 4 × 10 5

An experimental investigation of supercritical heat transfer in a three-rod bundle equipped with wire-wrap and grid spacers and cooled by carbon dioxide

Energy Technology Data Exchange (ETDEWEB)

Eter, Ahmad, E-mail: eng.eter@yahoo.com; Groeneveld, Dé, E-mail: degroeneveld@gmail.com; Tavoularis, Stavros, E-mail: stavros.tavoularis@uottawa.ca

2016-07-15

Highlights: • Heat transfer at supercritical pressures was studied experimentally in a three-rod bundle equipped with wire-wrap spacers or grid spacers. • Heat transfer deterioration occurred near the heated inlet under certain conditions. • Normal heat transfer was generally comparable to that in a tube and the predictions of a correlation. - Abstract: Heat transfer measurements in a three-rod bundle equipped with wire-wrap and grid spacers were obtained at supercritical pressures in the Supercritical University of Ottawa Loop (SCUOL). The tests were performed using carbon dioxide, as a surrogate fluid for water, flowing upwards for wide ranges of conditions, including conditions equivalent to the nominal and near-normal operating conditions of the proposed Canadian Super-Critical Water-Cooled Reactor. The test section contained three heated rods and three unheated rod segments with an outer diameter of 10 mm and a pitch-to-diameter ratio of 1.14; the heated length was 1500 mm. Detailed surface temperature measurements along and around the three heated rods were collected using internally traversed thermocouples. The following ranges of test conditions were covered, with equivalent water conditions given inside parentheses: pressure from 6.6 to 8.36 MPa (19.7–25 MPa); inlet temperature from 11 to 30 °C (330–371 °C); mass flux from 200 to 1175 kg m{sup −2} s{sup −1} (340–1822 kg m{sup −2} s{sup −1}); and wall heat flux from 1 to 175 kW m{sup −2} (11–1847 kW m{sup −2}). For one set of tests, the heated rods were fitted with a 1.3 mm OD wire wrap, having an axial pitch of 200 mm along the entire heated length; for a second set, the heated rods were fitted with grid spacers having a 5.3% flow blockage and located at 500 mm axial intervals. The effects of spacer configuration on heat transfer at supercritical pressures were documented and analyzed. The observed experimental trends were compared to those obtained in a experiment in a heated

Effects of spacers on blockage of coolant channels in clad melting accidents

Energy Technology Data Exchange (ETDEWEB)

Eggen, D. T.; Scale, T.; Hsieh, S. [Northwestern Univ., Evanston, IL (United States). The Technological Inst.

1977-07-01

The elements and configuration of these assemblies are representative of the current design for a GCFR. The fuel elements are stainless-steel clad, mixed-oxide spaced by a grid structure on 250 mm centers with a pitch of 9.5 mm, diameter, 7.2 mm, and cladding thickness, 0.5 m. Three series of experiments have been conducted to study the flow and disposition of molten cladding metal into a lower powered blanket region of the reactor following a loss of flow situation. The first two series used a simulant fuel-element bundle to simplify the experimental procedure and make visual observation possible. The 'fuel' was simulated by mullite rods 6.4 mm in diameter and 610 mm long. These were clad with a 50 Pb/50 Sn alloy tubing which was drawn onto the 'fuel'. The first series used cast spacers with webs of about 0.5-0.55 mm thickness placed 175 and 425 mm from the top end of the assembly. The second series used grid spacers fabricated of 0.25 mm alloy strips. This provided a more accurate representation of the hydraulic diameter. The bundle was encased in a hexagonal glass tube. The bundle was at 22/sup 0/C and the molten alloy was poured at a temperature of 260/sup 0/C (35/sup 0/C superheat). Motion pictures recorded the experiments and the bundle was sectioned for observation. The third set of experiments was done with a stainless steel bundle of 37 elements fabricated of mullite rods, 7.14 mm diameter. The stainless steel cladding had an O.D. of 8.41 mm. The element pitch was 11.1 mm. The grid spacers were prototypic. The experiment was conducted in an inert-gas tube furnace. The 'core fuel' cladding was melted in an induction furnace and the molten liquid flowed through the center seven element channels. X-ray pictures were taken after the tests and the bundle was sectioned for further study.

Construction method of pre assembled unit of bolt sphere grid

Science.gov (United States)

Hu, L. W.; Guo, F. L.; Wang, J. L.; Bu, F. M.

2018-03-01

The traditional construction of bolt sphere grid has many disadvantages, such as high cost, large amount of work at high altitude and long construction period, in order to make up for these shortcomings, in this paper, a new and applicable construction method is explored: setting up local scaffolding, installing the bolt sphere grid starting frame on the local scaffolding, then the pre assembled unit of bolt sphere grid is assembled on the ground, using small hoisting equipment to lift pre assembled unit to high altitude and install. Compared with the traditional installation method, the construction method has strong practicability and high economic efficiency, and has achieved good social and economic benefits.

Seismic behaviour of fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Song, Heuy Gap; Jhung, Myung Jo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1993-11-01

A general approach for the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced from earthquake. The dynamic responses such as fuel assembly shear force, bending moment and displacement, and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed. (Author) 9 refs., 24 figs., 1 tab.

Calculated and experimental research of WWER-1000 assembly vibration and fretting damage

International Nuclear Information System (INIS)

Drozdov, Y.; Afanasyev, A.; Makarov, V.; Tutnov, A.; Tutnov, A.; Alekseev, E.

2008-01-01

The report covers the methods and results of the latest analytical and experimental studies of fretting corrosion and natural vibrations of a WWER-1000 reactor fuel assemblies (FA). The process of fretting-corrosion was investigated using a multi-specimen facility that simulated fragments of fuel rod-to-spacer grid and lower support grid mating units. A computational model was developed for vibrations in the mechanical system of a fuel rod fragment and a spacer grid fragment. A calculational and experimental modal analysis of a FA was performed. Natural frequencies, modes and decrements of FA vibrations were determined and a satisfactory coincidence of analytical and experimental results was obtained. The assessment of fretting-corrosion process dynamics was made and its dependences on operational factors were obtained. (authors)

Spacer device for nuclear reactor fuel elements

International Nuclear Information System (INIS)

Anthony, A.J.; Gaines, A.L.; Krawiec, D.M.

1974-01-01

The grid-type spacer device consists of two rows of main spacers arranged parallel to each other with some space in between, the first row extending perpendicular to the second row. Parallel to the respective rows of main spacers there are rows of secondary spacers interlocked with the main spacers. The individual spacers are welded together at their points of intersection. A large number of spring cages are installed within the spacer device to hold in place the main spacers which are oriented at right angles relative to each other. In addition, the spring cages serve for supporting the fuel elements. The spacers are made of zirconium which does not greatly influence the neutron capture cross section of the reactor. The material of the spring cages with the spring elements is a nickel alloy. It has the necessary stress relaxation properties to be able to force the fuel elements against the spacers under the action of the spring. (DG) [de

Fuel assemblies for use in nuclear reactors

International Nuclear Information System (INIS)

Schluderberg, D.C.

1981-01-01

A fuel assembly for use in pressurized water cooled nuclear fast breeder reactors is described in which moderator to fuel ratios, conducive to a high Pu-U-D 2 O reactor breeding ratio, are obtained whilst at the same time ensuring accurate spacing of fuel pins without the parasitic losses associated with the use of spacer grids. (U.K.)

Preliminary CFD analysis methodology for flow in a LFR fuel assembly

International Nuclear Information System (INIS)

Catana, A.; Ioan, M.; Serbanel, M.

2013-01-01

In this paper a preliminary Computational Fluid Dynamics (CFD) analysis was performed in order to setup a methodology to be used for more complex coolant flow analysis inside ALFRED nuclear reactor fuel assembly. The core contains 171 separated fuel assembly, each consisting in a hexagonal array of 127 fuel rods. Three honey comb spacer grids are proposed along fuel rods with the aim to keep flow geometry intact during reactor operation. The main goal of this paper is to compute some hydraulic parameters: pressure, velocity, wall shear stress and turbulence parameters with and without spacer grids. In this analysis we consider an adiabatic case, so far no heat transfer is considered but we pave the road toward more complex thermo hydraulic analysis for ALFRED (LFR in general). The CAELINUX CFD distribution was used with its main components: Salome-Meca (for geometry and mesh) and Code-Saturne as mono-phase CFD solver. Paraview and Visist Postprocessors were used for data extraction and graphical displays. (authors)

Improvements in or relating to nuclear reactor fuel element assemblies

International Nuclear Information System (INIS)

Chetter, J.

1975-01-01

A description is given of a spacer grid comprising a substantially rigid grid structure formed from intersecting strip members defining cells which are penetrated by fuel pins bearing against rigid stops projecting inside the cells and spring locating members in the form of bow springs which extend longitudinally in the cells of the grid structure to hold the fuel pins against the rigid stops in the cells. The bow spring members of each line of cells extending in one direction across the grid structure have their corresponding ends interconnected by common longitudinal bridging strips to form a ladder spring assembly. (author)

CFD modeling of turbulent mixing through vertical pressure tube type boiling water reactor fuel rod bundles with spacer-grids

Science.gov (United States)

Verma, Shashi Kant; Sinha, S. L.; Chandraker, D. K.

2018-05-01

Numerical simulation has been carried out for the study of natural mixing of a Tracer (Passive scalar) to describe the development of turbulent diffusion in an injected sub-channel and, afterwards on, cross-mixing between adjacent sub-channels. In this investigation, post benchmark evaluation of the inter-subchannel mixing was initiated to test the ability of state-of-the-art Computational Fluid Dynamics (CFD) codes to numerically predict the important turbulence parameters downstream of a ring type spacer grid in a rod-bundle. A three-dimensional Computational Fluid Dynamics (CFD) tool (STAR-CCM+) was used to model the single phase flow through a 30° segment or 1/12th of the cross segment of a 54-rod bundle with a ring shaped spacer grid. Polyhedrons were used to discretize the computational domain, along with prismatic cells near the walls, with an overall mesh count of 5.2 M cell volumes. The Reynolds Stress Models (RSM) was tested because of RSM accounts for the turbulence anisotropy, to assess their capability in predicting the velocities as well as mass fraction of potassium nitrate measured in the experiment. In this way, the line probes are located in the different position of subchannels which could be used to characterize the progress of the mixing along the flow direction, and the degree of cross-mixing assessed using the quantity of tracer arriving in the neighbouring sub-channels. The predicted dimensionless mixing scalar along the length, however, was in good agreement with the measurements downstream of spacers.

Effect of top ligament blanking on reducing flow induced vibration of protective grid

Energy Technology Data Exchange (ETDEWEB)

Eom, Kyong Bo; Ryu, Joo Young; Kwon, Oh Joon; Park, Joon Kyoo; Jeon, Sang Youn; Suh, Jung Min [KEPCO NF Co., Daejeon (Korea, Republic of)

2012-10-15

The protective grid is a Inconel 718 spacer grid located just above the bottom nozzle in many kinds of fuel assemblies for PWR. The purpose of using protective grid is to capture debris before they flow up into the fuel assembly and get trapped by the other grids causing fuel rod damages as well as to provide support at the lower end plugs of fuel rods. Recently, it has been reported that strap failure has occurred in the protective grids and the flow induced vibration of the strap has resulted in the strap fatigue failure. After the root cause of the protective grid failure was found to be the flow induced vibration of the strap, KEPCO NF has made an effort to find the vibration tendencies of grid strap and draw vibration mitigation concepts of the protective grid strap. The vibration tendency and the effect of the vibration mitigation concept of the protective grid which have been found by the results of the loop tests and simulations in KEPCO NF are presented herein.

Surface-Tethered Iterative Carbohydrate Synthesis (STICS): A spacer study

Science.gov (United States)

Ganesh, N. Vijaya; Fujikawa, Kohki; Tan, Yih Horng; Nigudkar, Swati S.

2013-01-01

Comparative study of STICS using HPLC-assisted experimental set-up clearly demonstrated benefits of using longer spacer-anchoring systems. The use of mixed self-assembled monolayers helps to provide the required space for glycosylation reaction around the immobilized glycosyl acceptor. Both extension of the spacer length and using mixed self-assembled monolayers help to promote reaction and the beneficial effects may include moving the glycosyl acceptor further out into solution and providing additional conformational flexibility. It is possible that surface-immobilized glycosyl acceptors with a longer spacer (C8-O-C8)-lipoic acid have a higher tendency to mimic a solution-phase reaction environment than that of acceptors with shorter spacers. PMID:23822088

Investigation of a wire wrapped fuel assembly with the anisotropic Coarse-Grid-CFD (AP-CGCFD)

Energy Technology Data Exchange (ETDEWEB)

Viellieber, Mathias; Dietrich, Philipp; Class, Andreas [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). AREVA Nuclear Professional School (ANPS)

2013-07-01

Within this work we demonstrated the ability of the AP-CGCFD method to deal with complex geometries like wire wrapped spacer grid fuel assemblies. Both qualitative and quantitative values like the pressure profile and velocity structures could be reproduced from the detailed RANS CFD simulation. Furthermore we introduced a novel mathematical formulation of the method. Compared to state-of-the-art subchannel analyses, neither parameter tuning is needed, nor empirical or experimental input, to adjust the solvers for a specific geometry. Certainly, this method requires the user making educated decisions on the representative geometry segments and a suitable parameter space for the initial fine CFD simulations needed to extract the volumetric source terms. Since similar flow conditions repeat many times, the costs of the representative CFD simulations needed to extract the volumetric forces are much lower than a full simulation. Thus AP-CGCFD simulations are suitable for simulations of geometries where flow situations are repeating many times. (orig.)

Nuclear reactor fuel assembly grid

International Nuclear Information System (INIS)

Alder, J.L.; Kmonk, S.; Racki, F.R.

1981-01-01

A grid for a nuclear reactor fuel assembly which includes intersecting straps arranged to form a structure of egg crate configuration. The cells defined by the intersecting straps are adapted to contain axially extending fuel rods, each of which occupy one cell, while each control rod guide tube or thimble occupies the space of four cells. To effect attachment of each guide thimble to the grid, a short intermediate sleeve is brazed to the strap walls and the guide thimble is then inserted therein and mechanically secured to the sleeve walls. Each sleeve preferably, although not necessarily, is equipped with circumferentially spaced openings useful in adjusting dimples and springs in adjacent cells. To accurately orient each sleeve in position in the grid, the ends of straps extending in one direction project through transversely extending straps and terminate in the wall of the guide sleeve. Other straps positioned at right angles thereto terminate in that portion of the wall of a strap which lies next to a wall of the sleeve

Dependence of surface distribution of self-assembled InSb nanodots on surface morphology and spacer layer thickness

Energy Technology Data Exchange (ETDEWEB)

Godbole, M., E-mail: mohit.godbole@nmmu.ac.za [Department of Physics, Nelson Mandela Metropolitan University, PO Box 77000, Port Elizabeth 6031 (South Africa); Olivier, E.J. [Department of Physics, Nelson Mandela Metropolitan University, PO Box 77000, Port Elizabeth 6031 (South Africa); Coetsee, E.; Swart, H.C. [Department of Physics, University of the Free State, PO Box 339, Bloemfontein 9300 (South Africa); Neethling, J.H.; Botha, J.R. [Department of Physics, Nelson Mandela Metropolitan University, PO Box 77000, Port Elizabeth 6031 (South Africa)

2012-05-15

Self-assembled InSb nanodots (NDs) were grown on a GaSb (1 0 0) substrate using metal-organic vapour phase epitaxy (MOVPE). The effects of etching depth of the substrate and thickness of the GaSb buffer layer on the density and size distribution of single and double layer dots were studied for detector applications. The etch depth of the substrate was varied up to 30 {mu}m. In this particular study, the dots were grown at 450 Degree-Sign C and the GaSb spacer thickness was varied between 50 nm and 200 nm. The optimum substrate etch depth was found to be 30 {mu}m while the best spacer thickness was found to be 200 nm.

Dependence of surface distribution of self-assembled InSb nanodots on surface morphology and spacer layer thickness

International Nuclear Information System (INIS)

Godbole, M.; Olivier, E.J.; Coetsee, E.; Swart, H.C.; Neethling, J.H.; Botha, J.R.

2012-01-01

Self-assembled InSb nanodots (NDs) were grown on a GaSb (1 0 0) substrate using metal-organic vapour phase epitaxy (MOVPE). The effects of etching depth of the substrate and thickness of the GaSb buffer layer on the density and size distribution of single and double layer dots were studied for detector applications. The etch depth of the substrate was varied up to 30 μm. In this particular study, the dots were grown at 450 °C and the GaSb spacer thickness was varied between 50 nm and 200 nm. The optimum substrate etch depth was found to be 30 μm while the best spacer thickness was found to be 200 nm.

The AFA 3G fuel assembly: a proven design for high burnups

International Nuclear Information System (INIS)

Forat, C.; Florentin, F.

1999-01-01

The AFA 3G fuel assembly design is based on the wide experience gained with the AFA 2G fuel assembly. More than 9500 AFA 2G fuel assemblies have been loaded in different reactors, worldwide, reaching discharged burnups in the range of 45 - 55 GWd/tU. This experience confirmed the features of the AFA 2G, such as the grids and the vanes arrangement for thermal hydraulic performance, the concept of the fuel rod support within the grid which avoids any rod fretting or vibration phenomenon, the efficiency of the anti-debris device. The AFA 3G also relies on and benefits from the results of the world's largest R and D program, in-pile and out-of(pile testing by Framatome with EDF and CEA, with a special focus on corrosion-resistant fuel rod cladding. The AFA 3G exhibits the following enhancements: a reinforced structure, which improves resistance to assembly bow as well as its consequences in terms of RCCA insertion fuel handling and core physics obtained from: MONOBLOC TM guide thimbles, characterized by a thickened and enlarged tube and reinforced dash-pot; a hold down spring system which has been optimized to accommodate fuel assembly hydraulic lift-off forces and to meet the fuel assembly bow resistance requirement; widened recrystallized Zircaloy-4 spacer grids; a high resistance to corrosion due to the M5 TM Zirconium-Niobium-Oxygen alloy for the fuel rod cladding, which contributes also to the bow resistance of the fuel assembly; an enhanced thermal-hydraulic behavior promoted by well proven mixing vane array of AFA 2G spacer grids, combined with three additional Mid Span Mixing Grids; a very effective debris protection with the use of the TRAPPER TM bottom nozzle. With these improvements, the AFA 3G fuel assembly is able to reach discharge burnup of 60 GWd/tU with margins on important characteristics like corrosion behavior, assembly bow and thermal-hydraulic performance. The AFA 3G design is so convincing that major utilities have decided to shift their fuel

Grid structure for nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Wachter, W.J.; Akey, J.G.

1975-01-01

Described is a nuclear fuel element support system comprising an egg-crate-type grid made up of slotted vertical portions interconnected at right angles to each other, the vertical portions being interconnected by means of cross straps which are dimpled midway between their ends to engage fuel elements disposed within openings formed in the egg-crate assembly. The cross straps are disposed at an angle, other than a right angle, to the vertical portions of the assembly whereby their lengths are increased for a given span, and the total elastic deflection capability of the cell is increased. The assembly is particularly adapted for computer design and automated machine tool fabrication

Development of Mitsubishi high thermal performance grid 2 - overview of the development and Dnb test results

International Nuclear Information System (INIS)

Hoshi, M.; Imaizumi, M.; Mori, M.; Hori, K.; Ikeda, K.

2001-01-01

Spacer grid plays fundamental role in thermal performance of PWR fuel assembly. Grid spacer with higher thermal performance gives greater DNB (Departure from Nucleate Boiling) margin for the core. Mitsubishi has developed a prototype Zircaloy grid with higher thermal performance. In this paper, process of the development and DNB test results of the grid is presented. To achieve a goal to design grid with higher DNB performance, CFD (Computational Fluid Dynamics) and Freon DNB test are employed in the development. It is also concerned that the grid should be hydraulically compatible to existing grid. CFD is used in examining mixing capability and pressure drop for early stage of the development. Freon DNB test is used for preliminary checking of DNB performance for several design of the grids. After the final design is fixed, DNB test has been carried out at a high pressure / high temperature water test loop to verify the DNB performance. Also, hydraulic test has been done in a water test loop. The test results show that the grid has higher DNB performance and lower pressure loss coefficient compared with existing grid. It is also concluded that a combination of CFD and Freon DNB testing is successful tool for designing and development of grid. (authors)

Raman study of self-assembled InAs/InP quantum wire stacks with varying spacer thickness

Science.gov (United States)

Angelova, T.; Cros, A.; Cantarero, A.; Fuster, D.; González, Y.; González, L.

2008-08-01

Self-assembled InAs/InP (001) quantum wire stacks have been investigated by means of Raman scattering. The characteristics of the observed vibrational modes show clear evidence of confinement and atomic intermixing between As and P atoms from the wire and the spacer. The change in the intermixing with spacer layer thickness and growth temperature is investigated. Likewise, the effect of annealing on the exchange of As and P atoms is also studied. Resonance effects in confined and interface phonons are discussed for excitation in the vicinity of the InAs E1 critical point. Finally, the energy of the interface modes is related to the structural characteristics of the wires by comparing the experimental data with a lattice dynamic calculation based on the dielectric continuum model.

Mechanical characteristic evaluation of the mid grid spring in PWR fuel assembly

International Nuclear Information System (INIS)

Eum, K. B.; Lee, S. H.; Jeon, S. Y.; Kweon, Y. B.; Jeon, K. R.

2001-01-01

The spring load-deflection characteristic tests were performed for Westinghouse type 17x17 and 14x14 fuel assembly mid grids to evaluate the mechanical characteristics of the springs. Six kinds of prototype mid grids manufactured by KNFC were tested and two kinds of test methods were used: block test and in-grid test. The test results showed that all tested mid grid springs satisfied the criteria required at the beginning of fuel assembly life. In addition, the variation of spring characteristics resulting from the difference in the mechanical properties of spring material and spring shapes was investigated. And the validity of the test methods was discussed

Self-assembly of a [2 x 2] hydrogen bonded grid

NARCIS (Netherlands)

Lipkowski, P.R.; Bielejewska, A.G.; Kooijman, Huub; Spek, Anthony L.; Timmerman, P.; Reinhoudt, David

1999-01-01

Formation of 24 cooperative hydrogen bonds drives the spontaneous assembly of a rigid bifunctional trimelamine and bis(barbituric acid) to give selectively the [2 × 2] hydrogen-bonded grid, in preference to the corresponding [1 × 1] or polymeric assemblies.

CFD prediction of flow and phase distribution in fuel assemblies with spacers

Energy Technology Data Exchange (ETDEWEB)

Anglart, H.; Nylund, O. [ABB Atom AB, Vasteras (Switzerland); Kurul, N. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others

1995-09-01

This paper is concerned with the modeling and computation of multi-dimensional two-phase flows in BWR fuel assemblies. The modeling principles are presented based on using a two-fluid model in which lateral interfacial effects are accounted for. This model has been used to evaluate the velocity fields of both vapor and liquid phases, as well as phase distribution, between fuel elements in geometries similar to BWR fuel bundles. Furthermore, this model has been used to predict, in a detailed mechanistic manner, the effects of spacers on flow and phase distribution between, and pressure drop along, fuel elements. The related numerical simulations have been performed using a CFD computer code, CFDS-FLOW3D.

Evaluation of CASL boiling model for DNB performance in full scale 5x5 fuel bundle with spacer grids

Energy Technology Data Exchange (ETDEWEB)

Kim, Seung Jun [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2018-02-12

As one of main tasks for FY17 CASL-THM activity, Evaluation study on applicability of the CASL baseline boiling model for 5x5 DNB application is conducted and the predictive capability of the DNB analysis is reported here. While the baseline CASL-boiling model (GEN- 1A) approach has been successfully implemented and validated with a single pipe application in the previous year’s task, the extended DNB validation for realistic sub-channels with detailed spacer grid configurations are tasked in FY17. The focus area of the current study is to demonstrate the robustness and feasibility of the CASL baseline boiling model for DNB performance in a full 5x5 fuel bundle application. A quantitative evaluation of the DNB predictive capability is performed by comparing with corresponding experimental measurements (i.e. reference for the model validation). The reference data are provided from the Westinghouse Electricity Company (WEC). Two different grid configurations tested here include Non-Mixing Vane Grid (NMVG), and Mixing Vane Grid (MVG). Thorough validation studies with two sub-channel configurations are performed at a wide range of realistic PWR operational conditions.

3-D flow analyses for design of nuclear fuel spacer

Energy Technology Data Exchange (ETDEWEB)

Karouta, Z. [ABB Combustion Engineering, Windsor, CT (United States); GU, Chun-Yuan [ABB Corporate Research, Vaesteras (Sweden); Schoelin, B. [ABB Atom AB, Vaesteras (Sweden)

1995-09-01

The Computational Fluid Dynamics (CFD) code, CFDS-FLOW3D, was used to develop improved fuel designs for PWR cores. It was used primarily to understand the fluid dynamics of grid spacers, the mass transfer between subchannels caused by spacers and in the long term to develop two-phase models which enable prediction of critical heat flux in PWR fuel. A single subchannel of one grid span was modeled. In this model different spacer designs with mixing devices were analyzed. A special treatment of the boundary condition was developed making use of flow symmetry to model the mass transfer between different subchannels and minimize the size of the computational model. This reduced the computational model to a fraction of a subchannel using traditional periodic boundary conditions. The Navier-Stokes equation was solved for the liquid and the flow turbulence was modeled by k-{xi} turbulence model. The spacer and mixing device were treated as infinite thin surfaces in the model and a zero velocity condition and turbulent wall function were applied on each side of the thin surfaces. This approach simulated the swirl from the mixing devices well, but had the drawback of not predicting pressure drop accurately since the wake behind the plates and the acceleration effect of the spacers were ignored. CFDS-FLOW3D models with mixing devices were applied in the single-phase flow regime. Velocity profiles from the CFDS-FLOW3D models were compared to Laser Doppler Velocimeter measurements taken from the flow field downstream of spaces in a full scale, cold water test loop. The predicted axial and lateral velocity profiles were in good agreement with the measurements. The evaluation of the performance of different spacer devices was made by comparing the swirl ratio downstream of the grid spacers. It is planned to evaluate heat transfer coefficient downstream of the spaces, to implement two-phase flow models, and to model the superheated boundary layer on the surface of the fuel rod.

The influence of the preliminary garter spring spacer simulator clamping force in the pressure tube spacer -calandria tube hook-up simulator aging behaviour

International Nuclear Information System (INIS)

Gyongyosi, T.; Deloreanu, G.; Puiu, D.; Corbescu, B.; Anghel, N.; Dinu, E.

2016-01-01

The garter spring spacer is a specially constructed torsion spring used to fit-out the CANDU 6 fuel channel. The pressure tube ageing decreases the gap to the calandria tube. Continuous gap decrease directly affects the garter spring spacers behavior during fuel channel assembly operation. The preliminary clamping force value of the garter spring spacer assembly is important for its ageing behavior. This paper briefly describes the experimental technological facilities used for conducted the experiments and highlights some of the important moments during an experiment carried out in laboratory conditions, without using pressurized boiled water and irradiation working conditions. The results analysis and some conclusions are outlined at the end, pointing out that a garter spring spacer preliminary clamping force increase reduces the vibration response signal amplitude, and does not lead to its relaxation. The paper is dedicated to specialists working in research and technological engineering. (authors)

Improved spacers for high temperature gas-cooled heat exchangers

Energy Technology Data Exchange (ETDEWEB)

Nordstroem, L A [Swiss Federal Institute for Reactor Research, Wuerenlingen (Switzerland)

1984-07-01

Experimental and analytical investigations in the field of heat exchanger thermohydraulics have been performed at EIR for many years, Basic studies have been carried out on heat transfer and pressure loss for tube bundles of different geometries and tube surfaces. As a part of this overall R+D programme for heat exchangers, investigations have been carried out on spacer pressure loss in bundles with longitudinal flow. An analytical spacer pressure loss model was developed which could handle different types of subchannel within the bundle. The model has been evaluated against experiments, using about 25 spacers of widely differing geometries. In a gas-cooled reactor it is important to keep the pressure loss over the primary circuit heat exchangers to a minimum. In exchangers with grid spacers these contribute a significant proportion of the overall bundle losses. For example, in the HHT Recuperator, with a shell-side pressure loss of 3.5 % of the inlet pressure, the spacers cause about one half of this loss. Reducing the loss to, say, 2.5 % results in an overall increase in plant efficiency by more than 1 % - a significant improvement Preliminary analysis identified 5 geometries in particular which were chosen for experimental evaluation as part of a joint project with the SULZER Company, to develop a low pressure-loss spacer for HHT heat exchangers (longitudinal counter-flow He/He and He/H{sub 2}O designs). The aim of the tests was to verify the low pressure-loss characteristics of these spacer grid types, as well as the quality of the results calculated by the computer code analytical model. The experimental and analytical results are compared in this report.

Development of nuclear fuel for integrated reactor

Energy Technology Data Exchange (ETDEWEB)

Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M

1999-04-01

The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO{sub 2}-based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO{sub 2}-based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method.

Development of nuclear fuel for integrated reactor

International Nuclear Information System (INIS)

Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M.

1999-04-01

The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO 2 -based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO 2 -based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method

Process and device for fabricating nuclear fuel assembly grids

International Nuclear Information System (INIS)

Thiebaut, B.; Duthoo, D.; Germanaz, J.J.; Angilbert, B.

1991-01-01

The method for fabricating PWR fuel assembly grids consists to place the grid of which the constituent parts are held firmly in place within a frame into a sealed chamber full of inert gas. This chamber can rotate about an axis. The welding on one face at a time is carried out with a laser beam orthogonal to the axis orientation of the device. The laser source is outside of the chamber and the beam penetrates via a transparent view port

Analysis of dismantling possibility and unloading efforts of fuel assemblies from core of WWER

International Nuclear Information System (INIS)

Danilov, V.; Dobrov, V.; Semishkin, V.; Vasilchenko, I.

2006-01-01

The computation methods of optimal dismantling sequence of fuel assemblies (FA) from core of WWER after different operating periods and accident conditions are considered. The algorithms of fuel dismantling sequence are constructed both on the basis of analysis of mutual spacer grid overlaps of adjacent fuel assemblies and numerical structure analysis of efforts required for FA removal as FA heaving from the core. Computation results for core dismantling sequence after 3-year operating period and LB LOCA are presented in the paper

Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

Energy Technology Data Exchange (ETDEWEB)

Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong [Kookmin Univ., Seoul (Korea, Republic of)

2007-03-15

The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow.

Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

International Nuclear Information System (INIS)

Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong

2007-03-01

The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow

Nuclear reactor fuel assemblies and end fitting grid structures therefor

International Nuclear Information System (INIS)

Jabsen, F.S.

1978-01-01

An improved end fitting grid structure is described for nuclear fuel assemblies which overcomes the need for load-bearing control rod guide tubes and the expensive special fittings that these tubes required. (UK)

PLUS 7TM advanced fuel assembly development program for KSNPs and APR1400

International Nuclear Information System (INIS)

Kim, Kyutae; Stucker, David L.

2002-01-01

KNFC and Westinghouse have recently completed the development of the PLUS 7 TM advanced 16 X 16 fuel assembly for the Korean Standard Nuclear Plants (KSNPs) and the Advanced Power Reactor 1400 (APR 1400). This fuel design utilized the proven advanced design features including mixing vane spacer grids to increase critical heat flux performance, ZIRLO TM advanced materials to enable high-duty, high burnup fuel management and an optimized fuel rod diameter which improves fuel cycle cost while resulting in significant standardization of Korean fuel manufacture. PLUS 7 TM , also includes a patented spacer grid design with conformal fuel rod support designed to provide superior fuel rod wear/fretting resistance while minimizing pressure drop. This paper will present an overview of the PLUS 7 TM fuel assembly development process including a summary of the three-year design and testing program from a mechanical, neutronic, and thermal/hydraulic perspective. The PLUS 7 TM fuel for the KSNPs and the APR1400 reactors results in multi-million dollar per cycle savings in imported enriched uranium product for the Korean nuclear power program with technology specifically developed for Korea by experienced Korean engineers

Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

Directory of Open Access Journals (Sweden)

Waseem

2016-01-01

Full Text Available Fuel assembly of the PWR nuclear power plant is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the fuel assembly (FA of Chashma Nuclear Power Plant-1 (CHASNUPP-1 at room temperature in air. The non-linear contact and structural tensile analysis have been performed using ANSYS 13.0, in order to determine the fuel assembly (FA elongation behaviour as well as the location and values of the stress intensity and stresses developed in axial direction under applied tensile load of 9800 N or 2 g being the fuel assembly handling or lifting load [Y. Zhang et al., Fuel assembly design report, SNERDI, China, 1994]. The finite element (FE model comprises spacer grids, fuel rods, flexible contacts between the fuel rods and grid's supports system and guide thimbles with dash-pots and flow holes, in addition to the spot welds between spacer grids and guide thimbles, has been developed using Shell181, Conta174 and Targe170 elements. FA is a non-straight structure. The actual behavior of the geometry is non-linear due to its curvature or design tolerance. It has been observed that fuel assembly elongation values obtained through FE analysis and experiment [SNERDI Tech. Doc., Mechanical strength and calculation for fuel assembly, Technical Report, F3.2.1, China, 1994] under applied tensile load are comparable and show approximately linear behaviors. Therefore, it seems that the permanent elongation of fuel assembly may not occur at the specified load. Moreover, the values of stresses obtained at different locations of the fuel assembly are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Since the results of both studies (analytical and experimental are comparable, therefore, validation of the FE methodology is confirmed. The stress intensity of the FE model and maximum stresses developed along the guide thimbles in axial direction are

Cladding Heatup Prediction between Spacer Grids for the Downstream Effect Evaluation

International Nuclear Information System (INIS)

Park, J. Y.; Kim, M. W.

2009-01-01

Since a recirculation sump clogging issue by debris generated from high energy pipe line break had been invoked as GSI-191 in the US, many researches on this issue have been undertaken. Previous researches on this topic are well summarized in Bang et al. Due to comprehensive nature of the issue, it includes many area of research and one of them is the area of downstream effect evaluation. The downstream effect is involved with adverse effects of debris passing the sump screen on the downstream systems, components and piping including core and it can be further divided into an ex-vessel downstream effect and an in-vessel downstream effect. In the ex-vessel downstream effect, focus is laid on plugging of spray nozzle, wearing and abrasion of moving parts of pump and valve and etc. Otherwise, a debris effect on reactor core is focused in the in-vessel downstream effect. Since debris can be ingested in the core or the systems of downstream of sump screen during recirculation, basically the downstream effect influences long-term core cooling phase. With respect to the in-vessel downstream effect, an up-to-date evaluation methodology is well summarized in a topical report submitted to the US nuclear regulatory commission by the pressurized water reactor owners group (PWROG). The report evaluates various aspects of debris ingestion in the core such as blockage at the core inlet, collection of debris on fuel grids, plating-out of fuel, chemical precipitants, protective coatings effect and etc. Most of them are evaluated qualitative manner based on previous research results and geometrical consideration on fuel rod bundles but some of them are also backed up by quantitative calculations to corroborate the qualitative decisions. One of them is a cladding heatup calculation between spacer grids. This is done to demonstrate that the cladding temperature of a fuel rod between grids with debris deposited on the clad surface in a post- LOCA recirculation environment is below

Apparatus and method for loading fuel rods into grids of a fuel assembly

International Nuclear Information System (INIS)

De Mario, E.E.; Burman, D.L.; Olson, C.A.; Secker, J.R.

1987-01-01

This patent describes a fuel assembly having fuel rods and at least one grid formed of interleaved straps and yieldable springs, the interleaved straps defining hollow cells aligned in rows and columns thereof for receiving the respective fuel rods. A pair of the springs are disposed within each of the cells for engaging and supporting one of the fuel rods when received in the cell. An apparatus is described for facilitating the loading of the fuel rods into the grid of the fuel assembly, comprising: (a) first mean insertable concurrently into the cells of the grid for engaging and moving the springs from respective first positions in which each pair of springs will engage a respective fuel rod when disposed within the grid cell to respective second positions in which each pair of springs is disengaged from the respective fuel rod when disposed within the grid cell; (b) a pair of second means, one of the pair of the second means being insertable concurrently into the rows of the cells of the grid and the other of the pair of second means being insertable concurrently into the column of the cells

Spacing grid intended for nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Patterson, J.F.; Flora, B.S.

1977-01-01

This invention concerns a new improved type of spacing grid that can be used in nuclear reactor fuel assemblies. Under the invention a spacing grid is provided, preferably of the bimetallic type. This grid includes a set of flexible inconel strips positioned by structural 'zircalloy' fittings, having relatively low neutron absorption characteristics in comparison with systems where the flexible strips are welded in position, or where the spring forms an integral part of the structure. The openings for the fuel elements which are defined by the structural fittings intercrossing are fitted internally with bosses which work in conjunction with a spring directed downwards as from the flexible strip so as to position the individual fuel rods in their respective openings inside the grid structure. These flexible strips are arranged in rows extending in directions which depend on the particular design of the fuel asembly and which contain flexible components so distributed that the loads of the individual springs tend to equalize each other mutually. The reaction load exerting itself on the supporting structure is reduced to the minimum, and this results in a lesser distortion in the reactor and an equalisation of the spring loads [fr

The development of flow test technology for PWR fuel assembly

International Nuclear Information System (INIS)

Chung, Moon Ki; Cha, Chong Hee; Chung, Chang Hwan; Chun, Se Young; Song, Chul Hwa; Chung, Heung Joon; Won, Soon Yeun; Cho, Yeong Rho; Kim, Bok Deuk

1988-05-01

KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

Experimental and numerical investigation of water flow through spacer grids of nuclear fuel elements using the Open FOAM code; Investigação numérica e experimental do escoamento de água através de grades espaçadoras de elementos combustíveis nucleares utilizando o código OpenFOAM

Energy Technology Data Exchange (ETDEWEB)

Vidal, Guilherme A.M.; Vieira, Tiago A.S.; Castro, Higor F.P., E-mail: gvidal.ufmg@gmail.com, E-mail: tiago.vieira.eng@gmail.com, E-mail: higorfabiano@hotmail.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Mecânica; Santos, André A.C. dos; Silva, Vitor V. A.; Barros Filho, José A., E-mail: aacs@cdtn.br, E-mail: vitors@cdtn.br, E-mail: jabf@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

With the advancement and development of computational tools, the studies of thermofluidodynamic behavior in nuclear fuel elements have been developed in recent years. Of the devices present in these elements, the spacing grids received more attention. They have kept the fuel rods equally spaced and have fins that aim to improve the heat transfer process between the water and the fuel element. Therefore, the grids present an important structural and thermal function. This work was carried out with the purpose of verifying and validating simulations of spacer grids using OpenFOAM (2017) software of Computational Fluid Dynamics (CFD). The simulations were validated using results obtained through the commercial CFD program, Ansys CFX, and experiments available in the literature and obtained in test sections assembled on the Water-Air Circuit (CCA) of the CDTN thermo-hydraulic laboratory.

Investigation of grid-enhanced two-phase convective heat transfer in the dispersed flow film boiling regime

International Nuclear Information System (INIS)

Miller, D.J.; Cheung, F.B.; Bajorek, S.M.

2013-01-01

Highlights: • Experiments were done in the RBHT facility to study the droplet flow in rod bundle. • The presence of a droplet field was found to greatly enhance heat transfer. • A second-stage augmentation was observed downstream of a spacer grid. • This augmentation is due to the breakup of liquid ligaments downstream of the grid. - Abstract: A two-phase dispersed droplet flow investigation of the grid-enhanced heat transfer augmentation has been done using steam cooling with droplet injection experimental data obtained from the Penn State/NRC Rod Bundle Heat Transfer (RBHT) facility. The RBHT facility is a vertical, full length, 7 × 7-rod bundle heat transfer facility having 45 electrically heated fuel rod simulators of 9.5 mm (0.374-in.) diameter on a 12.6 mm (0.496-in.) pitch which simulates a portion of a PWR fuel assembly. The facility operates at low pressure, up to 4 bars (60 psia) and has over 500 channels of instrumentation including heater rod thermocouples, spacer grid thermocouples, closely-spaced differential pressure cells along the test section, several fluid temperature measurements within the rod bundle flow area, inlet and exit flows, absolute pressure, and the bundle power. A series of carefully controlled and well instrumented steam cooling with droplet injection experiments were performed over a range of Reynolds numbers and droplet injection flow rates. The experimental results were analyzed to obtain the axial variation of the local heat transfer coefficients along the rod bundle. At the spacer grid location, the flow was found to be substantially disrupted, with the hydrodynamic and thermal boundary layers undergoing redevelopment. Owing to this flow restructuring, the heat transfer downstream of a grid spacer was found to be augmented above the fully developed flow heat transfer as a result of flow disruption induced by the grid. Furthermore, the presence of a droplet field further enhanced the heat transfer as compared to single

Participation in benchmark MATIS-H of NEA/OCDE: uses CFD codes applied to nuclear safety. Study of the spacer grids in the fuel elements

International Nuclear Information System (INIS)

Pena-Monferrer, C.; Chiva, S.; Munoz-cobo, J. L.; Vela, E.

2012-01-01

This paper develops participation in benchmark MATIS-H, promoted by the NEA / OECD-KAERI, involving the study of turbulent flow in a rod beam with spacers in an experimental installation. Its aim is the analysis of hydraulic behavior of turbulent flow in the subchannels of the fuel elements, essential for the improvement of safety margins in normal and transient operations and to maximize the use of nuclear energy through an optimal design of grids.

A simplified approach to WWER-440 fuel assembly head benchmark

International Nuclear Information System (INIS)

Muehlbauer, P.

2010-01-01

The WWER-440 fuel assembly head benchmark was simulated with FLUENT 12 code as a first step of validation of the code for nuclear reactor safety analyses. Results of the benchmark together with comparison of results provided by other participants and results of measurement will be presented in another paper by benchmark organisers. This presentation is therefore focused on our approach to this simulation as illustrated on the case 323-34, which represents a peripheral assembly with five neighbours. All steps of the simulation and some lessons learned are described. Geometry of the computational region supplied as STEP file by organizers of the benchmark was first separated into two parts (inlet part with spacer grid, and the rest of assembly head) in order to keep the size of the computational mesh manageable with regard to the hardware available (HP Z800 workstation with Intel Zeon four-core CPU 3.2 GHz, 32 GB of RAM) and then further modified at places where shape of the geometry would probably lead to highly distorted cells. Both parts of the geometry were connected via boundary profile file generated at cross section, where effect of grid spacers is still felt but the effect of out flow boundary condition used in the computations of the inlet part of geometry is negligible. Computation proceeded in several steps: start with basic mesh, standard k-ε model of turbulence with standard wall functions and first order upwind numerical schemes; after convergence (scaled residuals lower than 10-3) and near-wall meshes local adaptation when needed, realizable k-ε of turbulence was used with second order upwind numerical schemes for momentum and energy equations. During iterations, area-average temperature of thermocouples and area-averaged outlet temperature which are the main figures of merit of the benchmark were also monitored. In this 'blind' phase of the benchmark, effect of spacers was neglected. After results of measurements are available, standard validation

LES analysis of the flow in a simplified PWR assembly with mixing grid

Science.gov (United States)

Bieder, Ulrich; Falk, Francois

2014-06-01

The flow in fuel assemblies of PWRs with mixing grids has been analyzed with CFD calculations by numerous authors. The comparisons between calculation and experiment are focused on the flow in the near wake of the mixing grid, i.e. on the flow in the first 10 to 20 hydraulic diameters (dh) downstream of the grid. In the study presented here, the comparison between the measurements in the AGATE facility (5x5 tube bundle) and TrioU calculations is done for the whole distance between two successive mixing grids that is up to 0.6m downstream of the grid. The AGATE experiments have originally not been designed for CFD validation but to characterize different types of mixing grids. Nevertheless, the quality of the experimental data allows the quantitative comparison between measurement and calculation. The conclusions of the comparison are summarized below: Linear turbulent viscosity models seem to work rather well as long as the cross flow velocity in the rod gaps is advection controlled, that is directly downstream of the mixing grid, Further downstream, when the cross flow velocity is reduced and isotropic turbulence becomes a more and more important mixing phenomena, linear viscosity models will fail, The mixing grid affects the cross flow velocity up to the successive grid at a distance of about 50dh. The flow in fuel assemblies is never similar to that in undisturbed rod bundles. The test section of the AGATE facility has been discretized on 300 million control volumes by using a staggered grid approach on tetrahedral meshes. 20 days of CPU on 4600 nodes of the HPC machine CURIE of the CCRT was necessary to calculate the statistics of the turbulent flow, in particular the mean velocity and the RMS of the turbulent fluctuations.

LES analysis of the flow in a simplified PWR assembly with mixing grid

International Nuclear Information System (INIS)

Bieder, Ulrich; Fauchet, Gauthier; Falk, Francois

2014-01-01

The flow in fuel assemblies of Pressurized Water Reactors (PWR) with mixing grids has been analysed with Computational Fluid Dynamics (CFD) by numerous authors. The comparisons between calculation and experiment are mostly focused on the flow in the near wake of the mixing grid, i.e. on the flow in the first 5 to 10 hydraulic diameters (dh) downstream of the grid. In the study presented here, the comparison between the measurements in the AGATE facility (5 * 5 tube bundle) and Trio-U calculations is done for the whole distance between two successive mixing grids that is up to about 50 d h downstream of the grid. The AGATE experiments have originally not been designed for CFD validation but to characterize different types of mixing grids. Nevertheless, the quality of the experimental data allows the quantitative comparison between measurement and calculation. The conclusions of the comparison are summarized below: Linear turbulent viscosity models seem to work rather well as long as the cross flow velocity in the rod gaps is advection controlled, that is directly downstream of the mixing grid, Further downstream, when the cross flow velocity is reduced and anisotropic turbulence becomes a more and more important mixing phenomena, linear viscosity models can fail, The mixing grid affects the cross flow velocity up to the successive grid. The flow in fuel assemblies is never similar to that in undisturbed rod bundles. The test section of the AGATE facility has been discretized on 300 million control volumes by using a staggered grid approach on tetrahedral meshes. 20 days of CPU on 4600 cores of the High Performance Computer (HPC) cluster CURIE of the Centre de Calcul, Recherche et Technologie (CCRT) were necessary to converge the statistics of the turbulent fluctuations, completely converge the mean velocity and incompletely converge the RMS of the turbulent fluctuations. (authors)

LES analysis of the flow in a simplified PWR assembly with mixing grid

International Nuclear Information System (INIS)

Bieder, U.; Falk, F.

2013-01-01

The flow in fuel assemblies of PWRs with mixing grids has been analyzed with CFD calculations by numerous authors. The comparisons between calculation and experiment are focused on the flow in the near wake of the mixing grid, i.e. on the flow in the first 10 to 20 hydraulic diameters (d h ) downstream of the grid. In the study presented here, the comparison between the measurements in the AGATE facility (5*5 tube bundle) and Trio U calculations is done for the whole distance between two successive mixing grids that is up to 0.6 m downstream of the grid. The AGATE experiments have originally not been designed for CFD validation but to characterize different types of mixing grids. Nevertheless, the quality of the experimental data allows the quantitative comparison between measurement and calculation. The conclusions of the comparison are summarized below. First, the linear turbulent viscosity models seem to work rather well as long as the cross flow velocity in the rod gaps is advection controlled, that is directly downstream of the mixing grid. Secondly, further downstream, when the cross flow velocity is reduced and isotropic turbulence becomes a more and more important mixing phenomena, linear viscosity models will fail. Thirdly, the mixing grid affects the cross flow velocity up to the successive grid at a distance of about 50 d h . The flow in fuel assemblies is never similar to that in undisturbed rod bundles. The test section of the AGATE facility has been discretized on 300 million control volumes by using a staggered grid approach on tetrahedral meshes. 20 days of CPU on 4600 nodes of the HPC machine CURIE of the CCRT (Computer Center for Research and Technology - France) was necessary to calculate the statistics of the turbulent flow, in particular the mean velocity and the RMS of the turbulent fluctuations. (authors)

Prediction of droplet deposition around BWR fuel spacer by FEM flow analysis

International Nuclear Information System (INIS)

Yamamoto, Yasushi; Morooka, Shinichi

1997-01-01

The critical power of the BWR fuel assembly has been remarkably increased. That increase mainly depends on the improvement of the spacer which keeps fixed gaps between fuel rods. So far, these improvements have been carried out on the basis of what developers consider to be appropriate and the results of mockup tests of the BWR fuel assembly. However, continued reliance on these approaches for the development of a higher performance fuel assembly will prove time-consuming and costly. Therefore, it is hoped that the spacer effects for the critical power can be investigated by computer simulation, and it is significantly important to develop the critical power prediction method. Direct calculation of the two-phase flow in a BWR fuel channel s still difficult. Accordingly, a new method for predicting the critical power was proposed. Our method consists of CFD (computer fluid dynamics) code based on the single-phase flow analysis method and the subchannel analysis code. To verify our method, the critical power predictions for various spacer geometries were performed. The predicted results of the critical power were compared with the experimental data. The result of the comparison showed a good agreement and the applicability of our method for various spacer geometries. (author)

Nondestructive examination of Oconee 1 fuel assemblies after four cycles of irradiation

International Nuclear Information System (INIS)

Pyecha, T.D.; Mayer, J.T.; Guthrie, B.A. III; Riordan, J.E.

1980-12-01

Five B and W Mark B (15 x 15) pressurized water reactor fuel assemblies were nondestructively examined after four cycles of irradiation in the Oconee 1 reactor. Four of the five assemblies examined had a burnup of 40,000 MWd/mtU; the fifth assembly had a burnup of 36,800 MWd/mtU. This effort is part of a Department of Energy program to improve uranium utilization by extending the burnup of light water reactor fuel. The examinations were conducted in the Oconee 1 and 2 spent fuel storage pool. Data obtained included fuel assembly and fuel rod dimensions, water channel spacings, spacer grid and holddown spring forces, fuel column stack and axial gap lengths, and crud samples. The results indicate that the assemblies performed well through four cycles of operation; all of the data were within design limits

Experimental determination of local temperature field variations due to spacer grids in the cladding tubes of a rod cluster flowed through by sodium

International Nuclear Information System (INIS)

Moeller, R.; Tschoeke, H.

1978-01-01

If spacer grids are used to keep the fuel rods in their places - as in the fuel elements of the SNR series, exact tests are necessary to find out whether and to what extent temperature peaks near the supporting points affect cladding tube design. To clarify this special problem, experimental investigations have been carried out for the first time in a rod cluster model of the SNR-300 fuel element cross-flowed with sodium. The investigations and findings so far are reported on. (orig./RW) [de

Efficient self-assembly of DNA-functionalized fluorophores and gold nanoparticles with DNA functionalized silicon surfaces: the effect of oligomer spacers

Science.gov (United States)

Milton, James A.; Patole, Samson; Yin, Huabing; Xiao, Qiang; Brown, Tom; Melvin, Tracy

2013-01-01

Although strategies for the immobilization of DNA oligonucleotides onto surfaces for bioanalytical and top-down bio-inspired nanobiofabrication approaches are well developed, the effect of introducing spacer molecules between the surface and the DNA oligonucleotide for the hybridization of nanoparticle–DNA conjugates has not been previously assessed in a quantitative manner. The hybridization efficiency of DNA oligonucleotides end-labelled with gold nanoparticles (1.4 or 10 nm diameter) with DNA sequences conjugated to silicon surfaces via hexaethylene glycol phosphate diester oligomer spacers (0, 1, 2, 6 oligomers) was found to be independent of spacer length. To quantify both the density of DNA strands attached to the surfaces and hybridization with the surface-attached DNA, new methodologies have been developed. Firstly, a simple approach based on fluorescence has been developed for determination of the immobilization density of DNA oligonucleotides. Secondly, an approach using mass spectrometry has been created to establish (i) the mean number of DNA oligonucleotides attached to the gold nanoparticles and (ii) the hybridization density of nanoparticle–oligonucleotide conjugates with the silicon surface–attached complementary sequence. These methods and results will be useful for application with nanosensors, the self-assembly of nanoelectronic devices and the attachment of nanoparticles to biomolecules for single-molecule biophysical studies. PMID:23361467

Spacers for fuel rod clusters

International Nuclear Information System (INIS)

Jabsen, F.S.

1978-01-01

The proposition deals with the fixing of nuclear fuel element rods in a grid which consists of a number of crossed Zy-plates which form cells. The rectangular cells have projections which serve as spacers for the fuel rods. According to the invention there are additional butt straps which can be moved in such a way that insertion and extraction of the fuel rods can be done without obstruction and they can be spring-loaded hold in their final position. (UWI) [de

Estimation of the nuclear fuel assembly eigenfrequencies in the probability sense

Directory of Open Access Journals (Sweden)

Zeman V.

2014-12-01

Full Text Available The paper deals with upper and lower limits estimation of the nuclear fuel assembly eigenfrequencies, whose design and operation parameters are random variables. Each parameter is defined by its mean value and standard deviation or by a range of values. The gradient and three sigma criterion approach is applied to the calculation of the upper and lower limits of fuel assembly eigenfrequencies in the probability sense. Presented analytical approach used for the calculation of eigenfrequencies sensitivity is based on the modal synthesis method and the fuel assembly decomposition into six identical revolved fuel rod segments, centre tube and load-bearing skeleton linked by spacer grids. The method is applied for the Russian TVSA-T fuel assembly in the WWER1000/320 type reactor core in the Czech nuclear power plant Temelín.

Technical verification for SMART fuel

International Nuclear Information System (INIS)

Chun, K. R.; Eom, K. B.; Seo, J. M.

2011-12-01

ο 1st year (2009.8.10-2009.12.31) - Test plan for candidated models/ - Fabrication of test components/ - Selection test and evaluation of test results/ - Fabrication of spacer grids for CHF test ο 2nd year (2010.1.1-2010.12.31) - Fabrication of spacer grids for CHF test/ - Fabrication of spacer grids for 17x17 ITL/ - Feasibility test and evaluation of test results/ - Fabrication of test CRA/ - Planning for out-of-pile test and fabrication of test components ο 3rd year (2011.1.1-2011.12.31) - Fabrication of out-of-pile test components and assemblies/ - Planning of components/fuel assemblies out-of-pile test/ - Performing components/fuel assemblies out-of-pile test/ - Test reports for out-of-pile tests/ - Licensing support

Fuel assembly for a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Ferrari, H M; Miller, D L; Tong, L S

1973-09-06

The subject of the patent is a spacer design applicable, primarily, to LWR, and especially, though not specifically PWR, fuel assemblies. The spacer consists of an egg-box type of assembly formed of interlocking pressed plates giving a square lattice whose openings accommodate fuel pins or regulating rods. The pressed plates are formed to provide pressed-out spring-like flanges which hold the fuel pins in position and guide the regulating rods. Additional pressed-out flanges ensure the correct configuration of the spacer structure. The spacer is designed to present as little resistance as possible to coolant flow.

Methodology development for estimating support behavior of spacer grid spring in core

International Nuclear Information System (INIS)

Yoon, Kyung Ho; Kang, Heung Seok; Kim, Hyung Kyu; Song, Kee Nam

1998-04-01

The fuel rod (FR) support behavior is changed during operation resulting from effects such as clad creep-down, spring force relaxation due to irradiation, and irradiation growth of spacer straps in accordance with time or increase of burnup. The FR support behavior is closely associated with time or increase of burnup. The FR support behavior is closely associated with FR damage due to fretting, therefore the analysis on the FR support behavior is normally required to minimize the damage. The characteristics of the parameters, which affect the FR support behavior, and the methodology developed for estimating the FR support behavior in the reactor core are described in this work. The FR support condition for the KOFA (KOrean Fuel Assembly) fuel has been analyzed by this method, and the results of the analysis show that the fuel failure due to the fuel rod fretting wear is closely related to the support behavior of FR in the core. Therefore, the present methodology for estimating the FR support condition seems to be useful for estimating the actual FR support condition. In addition, the optimization seems to be a reliable tool for establishing the optimal support condition on the basis of these results. (author). 15 refs., 3 tabs., 26 figs

Cellular Genetic Algorithm with Communicating Grids for Assembly Line Balancing Problems

Directory of Open Access Journals (Sweden)

BRUDARU, O.

2010-05-01

Full Text Available This paper presents a new approach with cellular multigrid genetic algorithms for the "I"-shaped and "U"-shaped assembly line balancing problems, including parallel workstations and compatibility constraints. First, a cellular hybrid genetic algorithm that uses a single grid is described. Appropriate operators for mutation, hypermutation, and crossover and two devoration techniques are proposed for creating and maintaining groups based on similarity. This monogrid algorithm is extended for handling many populations placed on different grids. In the multigrid version, the population of each grid is organized in clusters using the positional information of the chromosomes. A similarity preserving communication protocol between the clusters placed on different grids is introduced. The experimental evaluation shows that the multigrid cellular genetic algorithm with communicating grids is better than the hybrid genetic algorithm used for building it, whereas it dominates the monogrid version in all cases. Absolute performance is evaluated using classical benchmarks. The role of certain components of the cellular algorithm is explained and the effect of some parameters is evaluated.

Preliminary Study on the Fretting Wear Behaviors of a Duel Cooled Fuel Rod

Energy Technology Data Exchange (ETDEWEB)

Lee, Y.H.; Lee, K.H.; Kim, H.K. [KAERI, 150 Dukjin-dong Yuseon-gu Daejeon, 305-353 (Korea, Republic of)

2009-06-15

Based on MIT's concept, an innovative fuel development project was launched by KAERI that a substantial power up-rating could be realized by introducing an internally and externally double cooled annular fuel for current PWR reactors. In order to apply this duel cooled fuel to an OPR 1000 reactor system, geometrical features of structural parts in a fuel assembly should be changed except an overall dimension of a fuel assembly. Typical changes are summarized as fuel rod diameter and weight, shape and position of a spacer grid spring, etc. When considering a duel cooled fuel rod, its vibration characteristic and fretting behavior should be verified because the modified shape and dimension of spacer grid spring, fuel rod diameter and weight, number of spacer grid assembly are closely related to a flow-induced vibration in a duel cooled fuel assembly. In this study, based on FIV test results of 4x4 fuel assembly, fretting wear tests of an outer duel cooled fuel rod were performed by using an embossing type spacer grid spring that could adjust its spring stiffness. The discussion was focused on the evaluation of the optimized spring stiffness and spring position in 1x1 cell by analyzing the fretting wear results. (authors)

Experimental Simulation of Flow-Induced Vibration for Developing a Grid-to-Rod Fretting Model

Energy Technology Data Exchange (ETDEWEB)

Lee, Youngho; Kim, Hyungkyu; Kang, Heungseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

GTRF margin was calculated based on the fuel reliabilities program of operating power plants. But they have not accumulated sufficient experience under challenging operating conditions to be considered proven solutions. In addition, GTRF behaviors were significantly differed according to the plant types, operating condition and fuel types. So, analytical methods to resolve GTRF degradations are considered as difficult procedures for actual application. One of the most important problems is that it is difficult to evaluate the GTRF resistance of new spacer grid under operating power plant condition. Up to now, as a consequence, compliance with the fretting wear limit (typically 10% of the cladding thickness) is checked a posteriori, through post-irradiation examination. Therefore, in this study, rod simulation method for determining GTRF resistance of new spacer grid was proposed with a specially designed wear tester. This simulator enables us to examine the spacer grid shape effect under relatively short development period. In addition, for developing GTRF model, flow-induced vibration (FIV) was measured with different major variables such as GTR clearance, flow rate, etc. Fretting wear tests of nuclear fuel rods (i. e. grid-to-rod fretting) have been performed to examine the flow rate effect by using a specially designed test section with a simulated primary coolant. Based on above results, developed FIV-wear simulator could be effective to examine the spacer grid shape effect with short development period. Further study will be discussed on the GTR clearance effect with various spacer grid shapes.

Experimental study of mechanical properties on spacer in NHR

International Nuclear Information System (INIS)

Jiang Yueyuan; Shi Jibing; Xu Yong

2007-01-01

The spacer of NHR-200 is composed mainly of the inner, outer and cornual strips which are ranged in egg-crate of 12 x 12-3. First, the pre-distortion of three kinds of three-arc springs on reactor working condition and their related clipping-force ranges are analyzed in this paper. Secondly, the mechanical experiments of 1:1 prototype, such as the load-distortion experiments, which the load and distortion are respectively measured by strain gauge and displacement sensor, of three kinds of springs, rigid supports and the spacers in two different directions are carried out on a special experimental facility. The experimental results show that the spacer can completely meet the design demands of mechanical properties of the fuel assemblies in NHR-200. (authors)

Analytical model for calculation of the thermo hydraulic parameters in a fuel rod assembly

Energy Technology Data Exchange (ETDEWEB)

Cesna, B., E-mail: benas@mail.lei.l [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos g. 3, LT-44403 Kaunas (Lithuania)

2010-11-15

Research highlights: {yields} Proposed calculation model can be used for rapid calculation of the bundles with rods spaced by wire wrapping or honey type spacer grids. {yields} Model estimate three flow cross mixture mechanisms. {yields} Program DARS is enable to analyses experimental results. - Abstract: The paper presents the procedure of the cellular calculation of thermo hydraulic parameters of a single-phase gas flow in a fuel rod assembly. The procedure is implemented in the DARS program. The program is intended for calculation of the distribution of the gaseous coolant parameters and wall temperatures in case of arbitrary, geometrically specified, arrangement of the rods in fuel assembly and in case of arbitrary, functionally specified in space, heat release in the rods. In mathematical model the flow cross-section of the channel of intricate shape is conventionally divided to elementary cells formed by straight lines, which connect the centers of rods. Within the limits of a single cell the coolant parameters and the temperature of the corresponding part of the rod surface are assumed constant. The entire fuel assembly is viewed as a system of parallel interconnected channels. Program DARS is illustrated by calculation of a temperature mode of 85-rod assembly with spacers of wire wrapping on the rods.

A study of the friction and wear processes of the structural components of fuel assemblies for water-cooled and water moderated power reactors

International Nuclear Information System (INIS)

Makarov, V.; Afanasiev, A.; Matvienko, I.; Drozdov, Y.; Puchkov, V.

2011-01-01

The friction forces affect the fuel assembly (FA) strength at all the stages of its lifecycle. The paper covers the methods and the results of the pre-irradiation experimental studies of the static and dynamic processes the friction forces are involved in. These comprise the FA assembling at the manufacturer, fuel rod flow-induced vibration and fretting-wear in the fuel rod-to-cell friction pairs, rod cluster control assembly (RCCA) movement in the FA guide tubes, FA bowing, FA loading-unloading into the core, irradiation-induced growth and thermal-mechanical fuel rod-to-spacer grid interaction. (authors)

Reactor core and control rod assembly in FBR type reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

1993-01-01

Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

CFD method research on characteristics cells in rod bundle fuel assembly

International Nuclear Information System (INIS)

Chen Jie; Chen Bingyan; Zhang Hong

2011-01-01

Two characteristic cells are in AFA-3G fuel assembly, that is typical cell and control rod guide cell. And there are some rules on the arrangement of mixing vanes. For the two characteristic cells, mixing capability is evaluated axially from the point of the first and second kind of sub-channel with CFD method. Mass mixing and heat mixing are interaction but different with each other. Although the mass mixing in the first kind of sub-channel is stronger, the thermal capability of the two is to some tune from the point of heat transfer. In the experiment research on thermal-hydraulic performance of AFA-3G fuel assembly, the arrangements of mixing vanes should refer to the two spacer grids of characteristic cells. (authors)

Fuel-assembly vibration-induced neutron noise in PWRs

International Nuclear Information System (INIS)

Sweeney, F.J.; Renier, J.P.

1983-01-01

Space-dependent reactor kinetics calculations were performed to interpret observed increases in the amplitude of pressurized water reactor (PWR), ex-core neutron detector noise with increasing fuel burnup and correspondingly decreasing soluble boron concentration. These noise amplitude increases have occurred at both low frequencies (< 1.0 Hz) and in the 2.0- to 4.0-Hz frequency range. The noise amplitude increases in the 2.0- to 4.0-Hz frequency range have usually been accompanied by a decrease in the fundamental mode fuel assembly resonant frequency from 3.5 to 2.5 Hz over a fuel cycle, which has also been attributed to grid spacer spring relaxation

Device for a nuclear reactor. [Fuel element spacers

Energy Technology Data Exchange (ETDEWEB)

Foulds, R B; Kasberg, A H; Puechl, K H; Bleiberg, M L

1972-03-08

A spacer design for fuel element clusters for PWR type reactors is described. It consists of a frame supporting an egg-carton like grid each sector of which is provided with springs which grip the fuel pins. The spring design is such as to prevent fuel pin vibrations and at same time accommodate fuel pin deformations. Formulae for the calculation of natural frequencies, spring stiffness and friction loads are presented.

Fuel assembly

International Nuclear Information System (INIS)

Nomata, Terumitsu.

1993-01-01

Among fuel pellets to be loaded to fuel cans of a fuel assembly, fuel pellets having a small thermal power are charged in a region from the end of each of spacers up to about 50mm on the upstream of coolants that flow vertically at the periphery of fuel rods. Coolants at the periphery of fuel rods are heated by the heat generation, to result in voids. However, since cooling effect on the upstream of the spacers is low due to influences of the spacers. Further, since the fuel pellets disposed in the upstream region have small thermal power, a void coefficient is not increased. Even if a thermal power exceeding cooling performance should be generated, there is no worry of causing burnout in the upstream region. Even if burnout should be caused, safety margin and reliability relative to burnout are improved, to increase an allowable thermal power, thereby enabling to improve integrity and reliability of fuel rods and fuel assemblies. (N.H.)

The computer program ELCOM in the planning and structural analysis of PWR fuel elements: an example

International Nuclear Information System (INIS)

Silva Macedo, L.V. da

1990-01-01

Is's presented some results obtained with the ELCOM computer code, such as deflections, moments and natural frequencies, used in the design and structural analysis of PWR fuels assemblies. It's studied the behavior of these results varying the number of spacer grids, the rigidity of the joint between the fuel pin and the spacer grid, and the fuel assembly's boundary condition, considered in the analysis, in it's mounting into the core (if clamped-clamped, clamped-hinged or hinged-hinged). (author)

Critical heat flux near the critical pressure in heater rod bundle cooled by R-134A fluid: Effects of unheated rods and spacer grid

International Nuclear Information System (INIS)

Chun, Se-Y.; Shin, C.W.; Hong, S. D.; Moon, S. K.

2007-01-01

the case of the DNB (departure from nucleate boiling) at normal pressure conditions. The existence of a threshold pressure at which the CHF phenomenon disappears has been observed near the critical pressure. In the region where the pressure passes across the threshold pressure, CHF does not occur and the wall temperature variations increase monotonously according to the power level applied to the heater rods. The effects of unheated rods and spacer grid with mixing vane on the critical power have been investigated. The effect of unheated rods in the rod bundle on the critical power becomes smaller as the pressure approaches the critical pressure, and when the pressure exceeds 3.9 MPa, the unheated rods have little effect on the critical power. In the case of the rod bundle with the mixing vane spacer grids, the critical power shows larger value compared to that for the spacer grids without mixing vane. This trend is kept up to the pressure of 4.0 MPa (P/Pc =9.85) very close to the critical pressure

Experimental study on local resistance of two-phase flow through spacer grid with rod bundle

International Nuclear Information System (INIS)

Yan Chaoxing; Yan Changqi; Sun Licheng; Tian Qiwei

2015-01-01

The experimental study on local resistance of single-phase and two-phase flows through a spacer grid in a vertical channel with 3 × 3 rod bundle was carried out under the normal temperature and pressure. For the case of single-phase flow, the liquid Reynolds number covered the range of 290-18 007. For the case of two-phase flow, the ranges of gas and liquid superficial velocities were 0.013-3.763 m/s and 0.076-1.792 m/s, respectively. A correlation for predicting local resistance of single-phase flow was given based on experimental results. Eight classical two-phase viscosity formulae for homogeneous model were evaluated against the experimental data of two-phase flow. The results show that Dukler model predicts the experimental data well in the range of Re 1 < 9000 while McAdams correlation is the best one for Re 1 ≥ 9000. For all experimental data, Dukler model provides the best prediction with the mean relative error of 29.03%. A new correlation is fitted for the range of Re 1 < 9000 by considering mass quality, two- phase Reynolds number and liquid and gas densities, resulting in a good agreement with the experimental data. (authors)

Experimental study of mixing in a square array rod bundle with grid spacer

International Nuclear Information System (INIS)

Zong Guifang; Cai Zuti; Zhang Demei

1989-01-01

This paper describes the experimental study of mixing in a full scale 15x15 square array rod bundle fuel assembly with 10 mm diameter and 13.3 mm pitch. The experiment was carried out in an open water loop, K 2 CrO 4 was used as tracer. Each subchannel was sampled at the open bundle outlet. Titration, spectrophotometry and fibreoptic methods were used to measure the concentration. The Reynolds numbers ranged from 2.12x10 4 to 4.37x10 4 . For the turbulent mixing of the bare rod bundle, the results of this study agreed with the formulas recommended by other authors. Both flow visualisation studies and the quantitative analysis indicated that flow scattering caused by the grid has a little effect on the mixing. The cause has been examined in this paper. (orig.)

Enhancement of heat transfer in HPLWR fuel assemblies

International Nuclear Information System (INIS)

Bastron, A.; Hofmeister, J.; Meyer, L.; Schulenberg, T.

2005-01-01

A study on different methods for enhancement of heat transfer in fuel assemblies for a High Performance Light Water Reactor has been performed to indicate the potential for a further increase of core outlet temperature at given cladding temperatures, or for reduction of peak cladding temperatures at the envisaged core outlet temperature. As a result, the introduction of an artificial surface roughness or the use of a staircase type grid spacer should increase the heat transfer coefficient of the coolant at the cladding surface by more than a factor of two, which will reduce the peak cladding temperature by at least 50 degC. The paper provides further details for realization of these measures. (author)

Grid-based lattice summation of electrostatic potentials by assembled rank-structured tensor approximation

Science.gov (United States)

Khoromskaia, Venera; Khoromskij, Boris N.

2014-12-01

Our recent method for low-rank tensor representation of sums of the arbitrarily positioned electrostatic potentials discretized on a 3D Cartesian grid reduces the 3D tensor summation to operations involving only 1D vectors however retaining the linear complexity scaling in the number of potentials. Here, we introduce and study a novel tensor approach for fast and accurate assembled summation of a large number of lattice-allocated potentials represented on 3D N × N × N grid with the computational requirements only weakly dependent on the number of summed potentials. It is based on the assembled low-rank canonical tensor representations of the collected potentials using pointwise sums of shifted canonical vectors representing the single generating function, say the Newton kernel. For a sum of electrostatic potentials over L × L × L lattice embedded in a box the required storage scales linearly in the 1D grid-size, O(N) , while the numerical cost is estimated by O(NL) . For periodic boundary conditions, the storage demand remains proportional to the 1D grid-size of a unit cell, n = N / L, while the numerical cost reduces to O(N) , that outperforms the FFT-based Ewald-type summation algorithms of complexity O(N3 log N) . The complexity in the grid parameter N can be reduced even to the logarithmic scale O(log N) by using data-sparse representation of canonical N-vectors via the quantics tensor approximation. For justification, we prove an upper bound on the quantics ranks for the canonical vectors in the overall lattice sum. The presented approach is beneficial in applications which require further functional calculus with the lattice potential, say, scalar product with a function, integration or differentiation, which can be performed easily in tensor arithmetics on large 3D grids with 1D cost. Numerical tests illustrate the performance of the tensor summation method and confirm the estimated bounds on the tensor ranks.

Holddown device for nuclear fuel assembly

International Nuclear Information System (INIS)

Anthony, A.J.

1978-01-01

An apparatus for preventing ''floating'' of nuclear-reactor fuel assemblies due to hydraulic forces is disclosed. The apparatus uses a holddown column made of the same material as the core barrel. The column is positioned in a center guide-tube location in the fuel assembly in such a manner as to enable it either to slide within the center guide tube or, if the center guide tube is replaced by the column, to slide through openings in the spacer grids. The lower end of the holddown column engages the lower end fitting of the fuel assembly, and the upper end of the column engages a flow plate to which holddown force is applied. As a consequence of this arrangement, holddown force is transmitted from the flow plate through the holddown column to the lower end fitting. Movement of the fuel assembly is thereby prevented without a compression load being applied to the fuel-assemb1ly structure. In addition, variations due to thermal expansion in the distance between the lower core plate and the upper core plate are largely made up for by corresponding variations in the holddown column because the holddown column and the core barrel can be made of the same material

Impact forces on a core shroud of an excited PWR fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Collard, B.; Vallory, J. [CEA Cadarache, 13 - Saint Paul lez Durance (France)

2001-07-01

Seismic excitation of PWR internals may induce large motions of the fuel assemblies (FA). This could result in impact between assemblies or between assemblies and core shroud. Forces generated during these shocks are often the basis for the maximum design loads of the spacer grids and fuel rods. An experimental program has been conducted at the French Nuclear Reactor Directorate (CEA) to measure the impact forces of a reduced scale FA on the test section under different environmental conditions. Within the framework of the tests presented, the effect of the FA environment (air, stagnant water, water under flow) on the maximum impact forces measured at grid levels and on the energy dissipated during the shock is examined. A 'fluid cushioning' effect (dissipative) between the grids and the wall is sought. Experimental results show that the axial flow has a great influence on the impact forces. The greater the axial flow velocity is, the lower the impact forces are. The tests of impact of an assembly on a wall were analyzed compared to the tests carried out without impact. This analysis related on the measured forces of impact and the variation of the measured/computed total energy of the system. The whole of these tests in air and water shows that the 'fluid cushioning' effect required exists but is not significant. Thus the presence of water does not decrease the forces of impact, and does not amplify the quantity of energy dissipated during the shock. The fact that the 'fluid cushioning' effect is weak compared to more analytical tests probably comes from our 'not perfect' or 'realistic' conditions of tests which involve an angle between the grid and the wall at the shock moment.

Impact forces on a core shroud of an excited PWR fuel assembly

International Nuclear Information System (INIS)

Collard, B.; Vallory, J.

2001-01-01

Seismic excitation of PWR internals may induce large motions of the fuel assemblies (FA). This could result in impact between assemblies or between assemblies and core shroud. Forces generated during these shocks are often the basis for the maximum design loads of the spacer grids and fuel rods. An experimental program has been conducted at the French Nuclear Reactor Directorate (CEA) to measure the impact forces of a reduced scale FA on the test section under different environmental conditions. Within the framework of the tests presented, the effect of the FA environment (air, stagnant water, water under flow) on the maximum impact forces measured at grid levels and on the energy dissipated during the shock is examined. A 'fluid cushioning' effect (dissipative) between the grids and the wall is sought. Experimental results show that the axial flow has a great influence on the impact forces. The greater the axial flow velocity is, the lower the impact forces are. The tests of impact of an assembly on a wall were analyzed compared to the tests carried out without impact. This analysis related on the measured forces of impact and the variation of the measured/computed total energy of the system. The whole of these tests in air and water shows that the 'fluid cushioning' effect required exists but is not significant. Thus the presence of water does not decrease the forces of impact, and does not amplify the quantity of energy dissipated during the shock. The fact that the 'fluid cushioning' effect is weak compared to more analytical tests probably comes from our 'not perfect' or 'realistic' conditions of tests which involve an angle between the grid and the wall at the shock moment

Technical verification of advanced nuclear fuel for KSNPs

International Nuclear Information System (INIS)

Lee, C. B.; Bang, J. G.; Kim, D. H. and others

2002-03-01

KNFC has developed the advanced 16x16 fuel assembly for the Korean Standard Nuclear Plants through the three-year R and D project (from April 1999 to March 2002) under the Nuclear R and D program by MOST. The purpose of this project is to verify the advanced 16x16 fuel assembly for the Korean Standard Nuclear Plants being developed by KNFC during the same period. Verification tests for the advanced fuel assembly and its components such as characteristic test on the spacer grid spring and dimple, static buckling and dynamic impact test on the 5x5 partial spacer grid, the fuel rod vibration test supported by the PLUS7 mid-spacer grid, fretting wear test, turbulent flow structure test in wind tunnel and corrosion test were performed by using the KAERI facilities. Design reports and test results produced by KNFC were technically reviewed. For the domestic production of burnable poison rod, manufacturing technology of burnable poison pellets was developed

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)

Design of the Flow Plates for a Dual Cooled Fuel Assembly

International Nuclear Information System (INIS)

Kim, Jae Yong; Yoon, Kyung Ho; Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu

2009-01-01

In a dual cooled fuel assembly, the array and position of fuels are changed from those of a conventional PWR fuel assembly to achieve a power uprating. The flow plate provides flow holes to direct the heated coolant into/out of the fuel assembly and structural intensity to insure that the fuel rod is axially restrained within the spacer grids. So, flow plates of top/bottom end pieces (TEP/BEP) have to be modified into proper shape. Because the flow holes' area of a flow plate affects pressure drop, the flow holes' area must be larger than/equal to that of conventional flow plates. And design criterion of the TEP/BEP says that the flow plate should withstand a 22.241 kN axial load during handling lest a calculated stress intensity should exceed the Condition I allowable stress. In this paper, newly designed flow plates of a TEP/BEP are suggested and stress analysis is conducted to evaluate strength robustness of the flow plates for the dual cooled fuel assembly

Results of trial operation of the WWER advanced fuel assemblies

International Nuclear Information System (INIS)

Vasilchenko, I.; Dragunov, Y.; Mikhalchuk, A.

2001-01-01

The paper describes results from experimental operation of advanced WWER-1000 fuel assemblies (AFA) at five units in Balakovo NPP. Advanced fuel is developed according to the concept of standard WWER-1000 fuel assembly (jacket-free). The new features includes: 1) zirconium guiding channels (alloy E-635 and E-110) and spacer grids (alloy E-110); 2) integrated burnable absorber gadolinium; 3) extended service life of fuel assemblies (FA) and absorber rods (possibility of repair of FA); 4) improved adoption to reactor conditions. Some results of AFA pilot operation of a three year operation are presented and analyses of effectiveness of improvements are made concerning application of zirconium channels and grids; application of integrated burnable absorbers; extension of FA and absorbing rods service life and FA repairability. These new features of WWER-1000 fuel design allow: 1) to reduce the average fuel enrichment to the 3.77% instead of 4.31% in U-235; 2) to reduce the FA axial load in reactor hot state by 40%,; 3) increasing of fuel operation in reactor to the 30000 effective days with possibility to have a 5-year residence time in the reactor. The design of new generation FA for WWER-440 reactors involves few key changes. Fuel inventory in new fuel design is increased due to elongation of fuel stack and reducing the diameter of the central hole. Vibration stability is enhanced as a result of: no-play junction of the fuel rod with the lower grid; change of SG arrangements; strengthening of the lower grid unit; secure of the central tube in the gap. Water-uranium ration is increased. Introduction of all these kinds of modernization in a 5-year fuel cycle reduces fuel component in the energy cost to the 7%

CTF/STAR-CD off-line coupling for simulation of crossflow caused by mixing vane spacers in rod bundles

International Nuclear Information System (INIS)

Avramova, Maria

2011-01-01

Understanding the impact of the spacer grids on the reactor core thermal-hydraulics involves experimental mockup tests, numerical simulations, and development of reliable empirical or semi-empirical models. The state-of-the-art in modeling spacer effects on the thermal-hydraulic performance of the flow in Light Water Reactor (LWR) rod bundles employs numerical experiments by means of Computational Fluid Dynamics (CFD) calculations. The capabilities of the CFD codes are usually being validated against mock-up tests. Once validated, the CFD predictions can be used for improvement and development of more sophisticated models of the subchannel codes. Because of the involved computational cost, CFD codes can not be yet efficiently utilized for full bundle predictions, while advanced subchannel codes are a powerful tool for LWR safety and design analyses. Subchannel analyses are used for whole LWR core evaluations with relatively short CPU times and reasonable computer resources. The objectives of the presented work were to develop, implement, and qualify an innovative spacer grid model utilizing the Computational Fluid Dynamics within a framework of an efficient subchannel analysis tool. A methodology was developed for off-line coupling between the CFD code STAR-CD and the subchannel code CTF. The developed coupling scheme is flexible in axial mesh overlays. It was developed to be easily adapted to any pair of a CFD and a subchannel code. Separate modeling of the spacer grid effects on the diffusive and on the convective processes was implemented and successfully validated against experimental data. (author)

Forming Limit Diagrams of Zircaloy-4 and Zirlo Sheets for Stamping of Spacer Grids of Nuclear Fuel Rods

International Nuclear Information System (INIS)

Seo, Yun Mi; Hyun, Hong Chul; Lee, Hyung Yil; Kim, Nak Soo

2011-01-01

In this work, we investigated the theoretical forming limit models for Zircaloy-4 and Zirlo used for spacer grid of nuclear fuel rods. Tensile and anisotropy tests were performed to obtain stress-strain curves and anisotropic coefficients. The experimental forming limit diagrams (FLD) for two materials were obtained by dome stretching tests following NUMISHEET 96. Theoretical FLD depends on FL models and yield criteria. To obtain the right hand side (RHS) of FLD, we applied the FL models (Swift's diffuse necking, M-K theory, S-R vertex theory) to Zircaloy-4 and Zirlo sheets. Hill's local necking theory was adopted for the left hand side (LHS) of FLD. To consider the anisotropy of sheets, the yield criteria of Hill and Hosford were applied. Comparing the predicted curves with the experimental data, we found that the RHS of FLD for Zircaloy-4 can be described by the Swift model (with the Hill's criterion), while the LHS of the FLD can be explained by Hill model. The FLD for Zirlo can be explained by the S-R model and the Hosford's criterion (a = 8)

About calculation results of heat transfer in the fuel assembly clusters cooled by water with supercritical parameters

International Nuclear Information System (INIS)

Grabezhnaya, V.A.

2008-01-01

Paper reviews the numerical investigation into the heat transfer in the supercritical water cooled fuel assemblies on the basis of the various commercial codes. The turbulence available models specified in the codes describe adequately the experimental data in tubes within the range of flow temperatures away from the pseudocritical point, as well as under high mass velocities. There are k-ε type turbulence models that show qualitatively the local acceleration (slowdown) of the heat transfer in tubes, but they fail to describe the mentioned phenomena quantitatively. To determine the effect of grid spacers on the suppression of the heat transfer local slowdown and on the heat transfer acceleration in fuel assemblies and to ensure more accurate calculation of the fuel element cladding maximum temperature one should perform a number of the experiments making use of the fuel assembly models [ru

Spring retainer apparatus and method for facilitating loading of fuel rods into a fuel assembly grid

International Nuclear Information System (INIS)

De Mario, E.E.

1988-01-01

For use with a fuel assembly having at least one grid formed of interleaved straps defining hollow cells for respectively receiving fuel rods, at least some of the straps being disposed in pairs thereof so as to form springs in pairs therof being positioned in back-to-back relationships between adjacent ones of the cells, the springs in each pair thereof being configured to normally assume expanded positions in which they are displaced away from one another to engage fuel rods received in the respective cells and being deflectible to retracted positions in which they are displaced toward one another to allow loading of the fuel rods in the respective cells without engaging the springs, a spring retainer apparatus for facilitating the loading of the fuel rods into the cells of the fuel assembly grid is described comprising: (a) elongated holder bars, each holder bar being alignable with one of the pairs of the straps of the grid which defines the pairs of springs and extendible along, and in spaced relation from, the one strap pair and between and spaced from positions occupied by fuel rods when received in the cells of the grid; and (b) supported by each of the holder bars corresponding to the pairs of springs defined by the pair of straps aligned with the holder bar

Fuel assembly

International Nuclear Information System (INIS)

Kawai, Mitsuo.

1988-01-01

Purpose: To reduce the corrosion rate and suppress the increase of radioactive corrosion products in reactor water of nuclear fuel assemblies for use in BWR type reactors having spacer springs made of nickel based deposition reinforced type alloys. Constitution: Spacer rings made of nickel based deposition reinforced type alloy are incorporated and used as fuel assemblies after applying treatment of dipping and maintaining at high temperature water followed by heating in steams. Since this can remove the nickel leaching into reactor water at the initial stage, Co-58 as the radioactive corrosion products in the reactor water can be reduced, and the operation at in-service inspection or repairement can be facilitated to improve the working efficiency of the nuclear power plant. The dipping time is desirably more than 10 hours and more desirably more than 30 hours. (Horiuchi, T. )

Evaluation of a new welding process used to joint grids thimbles properly in 16 x 16 fuel assemblies, using the finite element method

Energy Technology Data Exchange (ETDEWEB)

Schettino, Carlos Frederico Mattos, E-mail: DPNcarlosschettino@inb.gov.b [Industrias Nucleares do Brasil S.A. (DPN/INB), Resende, RJ (Brazil). Diretoria de Producao Nuclear; Silva, Marcio Adriano Coelho da, E-mail: marcio.adriano@inb.gov.b [Industrias Nucleares do Brasil S.A. (GEACON/INB), Resende, RJ (Brazil). Gerencia de Analise Tecnica do Combustivel

2011-07-01

The present work aims to evaluate structurally the new welding process used to join the grids to the guide thimbles properly in 16 x 16 fuel assemblies. This new process is an increase of the number of welding points, 4 to 8, between grids and guide thimbles, giving more stiffness to the whole structure. A finite element model of the fuel assembly design was generated in the program ANSYS 12.1. To build this model were used elements BEAM-4 and several spring type elements. The analysis covered specific loads and displacements, simulating the boundaries conditions found during small deflection acting on the entire structure. The method used to development this analysis was the simulation of a finite element model performing a fuel assembly with four weld points on each grid cell containing the guide thimbles, and then the results of it was compare with another model, with eight weld points on each grid cell containing the guide thimbles. The behavior of the structure under the acting displacement and the related results of the analysis, mainly the stiffness, were satisfied. The results of this analysis were used to prove that the new grid to guide thimble welding process improve the dimensional stability when submitted to loads and displacements required on the fuel assembly design. The performed analysis provided INB to get more information of extreme importance, for the continuity of the development of new process of manufacturing and to improve the design of the current fuel assemblies used in reactors. (author)

Evaluation of a new welding process used to joint grids thimbles properly in 16 x 16 fuel assemblies, using the finite element method

International Nuclear Information System (INIS)

Schettino, Carlos Frederico Mattos; Silva, Marcio Adriano Coelho da

2011-01-01

The present work aims to evaluate structurally the new welding process used to join the grids to the guide thimbles properly in 16 x 16 fuel assemblies. This new process is an increase of the number of welding points, 4 to 8, between grids and guide thimbles, giving more stiffness to the whole structure. A finite element model of the fuel assembly design was generated in the program ANSYS 12.1. To build this model were used elements BEAM-4 and several spring type elements. The analysis covered specific loads and displacements, simulating the boundaries conditions found during small deflection acting on the entire structure. The method used to development this analysis was the simulation of a finite element model performing a fuel assembly with four weld points on each grid cell containing the guide thimbles, and then the results of it was compare with another model, with eight weld points on each grid cell containing the guide thimbles. The behavior of the structure under the acting displacement and the related results of the analysis, mainly the stiffness, were satisfied. The results of this analysis were used to prove that the new grid to guide thimble welding process improve the dimensional stability when submitted to loads and displacements required on the fuel assembly design. The performed analysis provided INB to get more information of extreme importance, for the continuity of the development of new process of manufacturing and to improve the design of the current fuel assemblies used in reactors. (author)

Three-dimensional analysis of the coolant flow characteristics in the fuel assemblies of VVER-1000 reactors

International Nuclear Information System (INIS)

Dinh Van Thin; Tran Thi Nhan

2015-01-01

Computational Fluid Dynamics (CFD) is a widely used method around the world for complex flow and heat industrial problems. In this paper, the coolant flow parameters were investigated in subchannels of VVER-1000 reactor’s fuel assemblies by ANSYS V14.5 programme. The different mesh solutions and turbulence models were carried out to deal with the water flow problems such as velocity distribution, streamline, temperature and pressure change as well as the hydraulic resistances of the spacer grids. The obtained results are good agreement with the measured values and the published reports from other authors. (author)

Fuel assembly

International Nuclear Information System (INIS)

Bando, Masaru.

1993-01-01

As neutron irradiation progresses on a fuel assembly of an FBR type reactor, a strong force is exerted to cause ruptures if the arrangement of fuel elements is not displaced, whereas the fuel elements may be brought into direct contact with each other not by way of spacers to cause burning damages if the arrangement is displaced. In the present invention, the circumference of fuel elements arranged in a normal triangle lattice is surrounded by a wrapper tube having a hexagonal cross section, wire spacers are wound therearound, and deformable spacers are distributed to optional positions for fuel elements in the wrapper tube. Interaction between the fuel elements caused by irradiation is effectively absorbed, thereby enabling to delay the occurrence of the rupture and burning damages of the elements. (N.H.)

Stress Linearization and Strength Evaluation of the BEP's Flow Plates for a Dual Cooled Fuel Assembly

International Nuclear Information System (INIS)

Kim, Jae Yong; Yoon, Kyung Ho; Kang, Heung Seok; Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu

2009-01-01

A fuel assembly is composed of 5 major components, such as a top end piece (TEP), a bottom end piece (BEP), spacer grids (SGs), guide tubes (GTs) and an instrumentation tube (IT) and fuel rods (FRs). There are no ASME criteria about all components except for a TEP/BEP. The TEP/BEP should satisfy stress intensity limits in case of condition A and B of ASME, Section III, Division 1 . Subsection NB. In a dual cooled fuel assembly, the array and position of fuels are changed from those of a conventional PWR fuel assembly to achieve a power uprating. The flow plates of top/bottom end pieces (TEP/BEP) have to be modified into proper shape to provide flow holes to direct the heated coolant into/out of the fuel assembly but structural intensity of these plates within a 22.241 kN axial loading should satisfy Tresca stress limits in ASME code. In this paper, stress linearization procedure and strength evaluation of a newly designed BEP for the dual cooled fuel assembly are described

Optimized spacer layer thickness for plasmonic-induced enhancement of photocurrent in a-Si:H

Energy Technology Data Exchange (ETDEWEB)

Saleh, Z. M., E-mail: zaki.saleh@aauj.edu, E-mail: zakimsaleh@yahoo.com; Nasser, H.; Özkol, E.; Günöven, M.; Abak, K. [Middle East Technical University, Center for Solar Energy Research and Applications (GÜNAM) (Turkey); Canli, S. [Middle East Technical University, Central Laboratory (Turkey); Bek, A.; Turan, R. [Middle East Technical University, Center for Solar Energy Research and Applications (GÜNAM) (Turkey)

2015-10-15

Plasmonic interfaces consisting of silver nanoparticles of different sizes (50–100 nm) have been processed by the self-assembled dewetting technique and integrated to hydrogenated amorphous silicon (a-Si:H) using SiNx spacer layers to investigate the dependence of optical trapping enhancement on spacer layer thickness through the enhancements in photocurrent. Samples illuminated from the a-Si:H side exhibit a localized surface plasmon resonance (LSPR) that is red-shifted with the increasing particle size and broadened into the red with the increasing spacer layer thickness. The photocurrent measured in a-Si:H is not only consistent with the red-shift and broadening of the LSPR, but exhibits critical dependence on the spacer layer thickness also. The samples with plasmonic interfaces and a SiNx spacer layer exhibit appreciable enhancement of photocurrent compared with flat a-Si:H reference depending on the size of the Ag nanoparticle. Simulations conducted on one-dimensional square structures exhibit electric fields that are localized near the plasmonic structures but extend appreciably into the higher refractive index a-Si:H. These simulations produce a clear red-shift and broadening of extinction spectra for all spacer layer thicknesses and predict an enhancement in photocurrent in agreement with experimental results. The spectral dependence of photocurrent for six plasmonic interfaces with different Ag nanoparticle sizes and spacer layer thicknesses are correlated with the optical spectra and compared with the simulations to predict an optimal spacer layer thickness.

Mechanical characterization tests of a candidate skeleton for X-Gen fuel assembly

International Nuclear Information System (INIS)

Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Kang Hee; Kim, Jae Yong; Lee, Young Ho

2007-09-01

Since the KNFC (KEPCO Nuclear Fuel Co.) requested a mechanical characterization tests of a candidate skeleton for X-Gen fuel assembly (some welding locations of a center guide tube are free of welding compared with the PLUS7 case) were requested, transverse vibration and stiffness tests were carried out by using the FAMeCT. The major results are as follows. - Transverse vibration test There was no distinguishable discrepancy in the free vibration characteristics between the skeleton without welding at some locations of a center guide tube and that of original assembly (PLUS7; welded at every spacer grid locations). The natural frequencies were measured as 6.8 - 6.9 for the 1st mode; 17.7 - 18.3 for the 2nd mode; 30.2 - 31.2 for the 3rd mode; 50.4 - 52.1 Hz for the 4th mode. As a result, the difference in the vibration characteristics was extremely small regardless of the number of welding of a center guide tube. - Transverse bending test. The transverse bending test results of the X-Gen no. 2 were similar to those of the PLUS7 skeleton. The relationship between the force and displacement was found linear. 521 N was observed at the deflection of 30 mm, and the stiffness at the 6th grid location (load exerting location) was 17.4, 16.3 N/mm in the two consecutive tests. The displacements at the grid locations lower than the 6th grid were at bit smaller than those upper than that due to a comparatively higher rigidity

Development of Mitsubishi high thermal performance grid 1 - CFD applicability for thermal hydraulic design

International Nuclear Information System (INIS)

Ikeda, K.; Hoshi, M.

2001-01-01

Mitsubishi applied the Computational Fluid Dynamics (CFD) evaluation method for designing of the new lower pressure loss and higher DNB performance grid spacer. Reduction of pressure loss of the grid has been estimated by CFD. Also, CFD has been developed as a design tool to predict the coolant mixing ability of vane structures, that is to compare the relative peak spot temperatures around fuel rods at the same heat flux condition. These evaluations have been reflected to the new grid spacer design. The prototype grid was manufactured and some flow tests were performed to examine the thermal hydraulic performance, which were predicted by CFD. The experimental data of pressure loss was in good agreement with CFD prediction. The CFD prediction of flow behaviors at downstream of the mixing vanes was verified by detail cross-flow measurements at rod gaps by the rod LDV system. It is concluded that the applicability of the CFD evaluation method for the thermal hydraulic design of the grid is confirmed. (authors)

Amorphous filler metal foils for brazing zirconium grid plates

International Nuclear Information System (INIS)

Plyushchev, A.N.; Kalin, B.A.; Fedotov, V.T.; Sevryukov, O.N.; Mamedova, T.T.; Shestakov, E.F.; Timoshin, S.N.

2001-01-01

A new amorphous ribbon filler metal of Zr-5.5 Fe-2.5 Be-1.0 Nb-8.0 Cu-2.0 Sn-0.4 Cr (mass %) with the temperature of melting onset of 745-750 deg C is designed to braze spacer grids of zirconium base alloys. The brazing conditions (780-790 deg C, 40-45 s) are determined which provide minimal standing at temperatures above 700 deg C (∼ 1.5 min) for spacer grids. Mechanical tests show that tensile strength of brazed joints is 55-59 kgf what is twice that of analogous welded joints. In addition, the brazed joints exhibit high corrosion resistance when testing in a distilled steam-water mixture at a temperature of 350 deg C and 16.5 MPa pressure for 10000 h [ru

Automatically closing swing gate closure assembly

Science.gov (United States)

Chang, Shih-Chih; Schuck, William J.; Gilmore, Richard F.

1988-01-01

A swing gate closure assembly for nuclear reactor tipoff assembly wherein the swing gate is cammed open by a fuel element or spacer but is reliably closed at a desired closing rate primarily by hydraulic forces in the absence of a fuel charge.

A functional analysis of the spacer of V(DJ recombination signal sequences.

Directory of Open Access Journals (Sweden)

Alfred Ian Lee

2003-10-01

Full Text Available During lymphocyte development, V(DJ recombination assembles antigen receptor genes from component V, D, and J gene segments. These gene segments are flanked by a recombination signal sequence (RSS, which serves as the binding site for the recombination machinery. The murine Jbeta2.6 gene segment is a recombinationally inactive pseudogene, but examination of its RSS reveals no obvious reason for its failure to recombine. Mutagenesis of the Jbeta2.6 RSS demonstrates that the sequences of the heptamer, nonamer, and spacer are all important. Strikingly, changes solely in the spacer sequence can result in dramatic differences in the level of recombination. The subsequent analysis of a library of more than 4,000 spacer variants revealed that spacer residues of particular functional importance are correlated with their degree of conservation. Biochemical assays indicate distinct cooperation between the spacer and heptamer/nonamer along each step of the reaction pathway. The results suggest that the spacer serves not only to ensure the appropriate distance between the heptamer and nonamer but also regulates RSS activity by providing additional RAG:RSS interaction surfaces. We conclude that while RSSs are defined by a "digital" requirement for absolutely conserved nucleotides, the quality of RSS function is determined in an "analog" manner by numerous complex interactions between the RAG proteins and the less-well conserved nucleotides in the heptamer, the nonamer, and, importantly, the spacer. Those modulatory effects are accurately predicted by a new computational algorithm for "RSS information content." The interplay between such binary and multiplicative modes of interactions provides a general model for analyzing protein-DNA interactions in various biological systems.

Computational comparison of the effect of mixing grids of 'swirler' and 'run-through' types on flow parameters and the behavior of steam phase in WWER fuel assemblies

International Nuclear Information System (INIS)

Shcherbakov, S.; Sergeev, V.

2011-01-01

The results obtained using the TURBOFLOW computer code are presented for the numerical calculations of space distributions of coolant flow, heating and boiling characteristics in WWER fuel assemblies with regard to the effect of mixing grids of 'Swirler' and 'Run-through' types installed in FA on the above processes. The nature of the effect of these grids on coolant flow was demonstrated to be different. Thus, the relaxation length of cross flows after passing a 'Run-through' grid is five times as compared to a 'Swirler'-type grid, which correlates well with the experimental data. At the same time, accelerations occurring in the flow downstream of a 'Swirler'-type grid are by an order of magnitude greater than those after a 'Run-through' grid. As a result, the efficiency of one-phase coolant mixing is much higher for the grids of 'Run-through' type, while the efficiency of steam removal from fuel surface is much higher for 'Swirler'-type grids. To achieve optimal removal of steam from fuel surface it has been proposed to install into fuel assemblies two 'Swirler'-type grids in tandem at a distance of about 10 cm from each other with flow swirling in opposite directions. 'Run-through' grids would be appropriate for use for mixing in fuel assemblies with a high non-uniformity of fuel-by-fuel power generation. (authors)

Nuclear fuel assemblies and fuel pins usable in such assemblies

International Nuclear Information System (INIS)

Jolly, R.

1982-01-01

A novel end cap for a nuclear fuel assembly is described in detail. It consists of a trisection arrangement which is received within a cell of a cellular grid. The cell contains abutment means with which the trisection comes into abutment. The grid also contains an abutment means for preventing the trisections from being inserted into the cell in an incorrect orientation. The present design allows fuel pins to be securely held in a hold-down grid of a sub-assembly. The design also allows easier dis-assembly of the swollen and embrittled fuel pins prior to reprocessing. (U.K.)

Fouling resistant membrane spacers

KAUST Repository

Ghaffour, Noreddine

2017-10-12

Disclosed herein are spacers having baffle designs and perforations for efficiently and effectively separating one or more membrane layers a membrane filtration system. The spacer (504) includes a body (524) formed at least in part by baffles (520) that are interconnected, and the baffles define boundaries of openings or apertures (525) through a thickness direction of the body of the spacer. Alternatively or additionally, passages or perforations (526A, 526B) may be present in the spacer layer or baffles for fluid flow there through, with the passages and baffles having a numerous different shapes and sizes.

Effects of the core grids on the burnout

International Nuclear Information System (INIS)

Katsaounis, A.; Fulfs, H.; Stein, M.

1977-01-01

This paper reports on burnout experiments carried out using freon 12 at four test sections with 6 x 6, 8 x 8, 7 x 7 rod bundles, and with annular geometries. The axial heat flux distribution of the heated rods is either uniform or simulating reactor conditions. For the rod bundle test sections original reactor grids of PWR type are used with mixing vanes or similar grids of PWR without vanes. For the annular test sections an orifice simulates the spacer. At all experiments without any expection the burnout occurs in front of a grid. The film boiling condition moves always from one front to the next front of a grid, but never just behind a grid. (orig./HP) [de

Construction for fissionable material

International Nuclear Information System (INIS)

Jabsen, F.S.

1978-01-01

This invention is directed toward a nuclear fuel assembly guide tube arrangement which restrains spacer grid movement due to coolant flow and which offers secondary means for supporting a fuel assembly during handling and transfer operations. (Auth.)

A seismic analysis of Korean standard PWR fuels under transition core conditions

International Nuclear Information System (INIS)

Kim, Hyeong Koo; Park, Nam Kyu; Jang, Young Ki; Kim, Jae Ik; Kim, Kyu Tae

2005-01-01

The PLUS7 fuel is developed to achieve higher thermal performance, burnup and more safety margin than the conventional fuel used in the Korean Standard Nuclear Plants (KSNPs) and to sustain structural integrity under increased seismic requirement in Korea. In this study, a series of seismic analysis have been performed in order to evaluate the structural integrity of fuel assemblies associated with seismic loads in the KSNPs under transition core conditions replacing the Guardian fuel, which is a resident fuel in the KSNP reactors, with the PLUS7 fuel. For the analysis, transition core seismic models have been developed, based on the possible fuel loading patterns. And the maximum impact forces on the spacer grid and various stresses acting on the fuel components have been evaluated and compared with the through-grid strength of spacer grids and the stress criteria specified in the ASME code for each fuel component, respectively. Then three noticeable parameters regarding as important parameters governing fuel assembly dynamic behavior are evaluated to clarify their effects on the fuel impact and stress response. As a result of the study, it has been confirmed that both the PLUS7 and the Guardian fuel sustain their structural integrity under the transition core condition. And when the damping ratio is constant, increasing the natural frequency of fuel assembly results in a decrease in impact force. The fuel assembly flexural stiffness has an effect increasing the stress of fuel assembly, but not the impact force. And the spacer grid stiffness is directly related with the impact force response. (author)

Structural Stator Spacers

DEFF Research Database (Denmark)

Rasmussen, Peter Omand; Andreasen, Jens H.; Pijanowski, J. M.

2001-01-01

This paper presents a powerful new design aspect to reduce acoustic noise and vibration of electro-magnetic origin for electrical machines, by introducing improved slot wedges referred to as "Structural Stator Spacers". These spacers, by using a very stiff dielectric and non magnetic material...

The Necessity of a New Type Test Rig for the Development of an Evaluation Method in Grid Fretting Problems

International Nuclear Information System (INIS)

Lee, Young-Ho; Kim, Hyung-Kyu

2007-01-01

A grid fretting problem is recognized as one of the most important degradation mechanisms even though the examination results of fretting experiments could be applied to the development and design of spacer grid structures. This is because it is difficult to develop a fretting wear model for a grid fretting problem due to the various wear mechanisms involved according to the mechanical and environmental variables, the contact condition with a spring/dimple and the material properties. A number of spring shapes has been developed in KAERI and their performance tests such as fretting wear, flow-induced vibration (FIV) tests, etc. have been carried out from a part unit to a full assembly scale. From the unit part fretting test results, one of the noticeable results is that the contacting force (normal load) was gradually decreased with increasing number of fretting cycles due to a depth increase and this behavior was closely related to the contacting spring shape. When considering the actual contact condition between a fuel rod and a spring/dimple, if a fretting wear progresses due to a FIV under a specific normal load exerted on the fuel rod by an elastic deformation of the spring, the contacting force between the fuel rod and dimple that are located in the opposite side should be decreased. Consequently, an evaluation of developed spacer grids against fretting wear damage should be performed with the results of 1x1 cell unit experiments because a contacting force is one of the most important variables that influences a fretting wear mechanism. The discussion was focused on the development procedure of a new test rig and its performance by using a 1x1 cell unit test rig. (authors)

A functional analysis of the spacer of V(D)J recombination signal sequences.

Science.gov (United States)

Lee, Alfred Ian; Fugmann, Sebastian D; Cowell, Lindsay G; Ptaszek, Leon M; Kelsoe, Garnett; Schatz, David G

2003-10-01

During lymphocyte development, V(D)J recombination assembles antigen receptor genes from component V, D, and J gene segments. These gene segments are flanked by a recombination signal sequence (RSS), which serves as the binding site for the recombination machinery. The murine Jbeta2.6 gene segment is a recombinationally inactive pseudogene, but examination of its RSS reveals no obvious reason for its failure to recombine. Mutagenesis of the Jbeta2.6 RSS demonstrates that the sequences of the heptamer, nonamer, and spacer are all important. Strikingly, changes solely in the spacer sequence can result in dramatic differences in the level of recombination. The subsequent analysis of a library of more than 4,000 spacer variants revealed that spacer residues of particular functional importance are correlated with their degree of conservation. Biochemical assays indicate distinct cooperation between the spacer and heptamer/nonamer along each step of the reaction pathway. The results suggest that the spacer serves not only to ensure the appropriate distance between the heptamer and nonamer but also regulates RSS activity by providing additional RAG:RSS interaction surfaces. We conclude that while RSSs are defined by a "digital" requirement for absolutely conserved nucleotides, the quality of RSS function is determined in an "analog" manner by numerous complex interactions between the RAG proteins and the less-well conserved nucleotides in the heptamer, the nonamer, and, importantly, the spacer. Those modulatory effects are accurately predicted by a new computational algorithm for "RSS information content." The interplay between such binary and multiplicative modes of interactions provides a general model for analyzing protein-DNA interactions in various biological systems.

Experimental determination of the local temperature distribution in the cladding tubes of a sodium-cooled pin bundle caused by grid spacers

International Nuclear Information System (INIS)

Moeller, R.; Tschoeke, H.

1980-01-01

The cladding tubes of reactor core elements are highly stressed structural elements. Their careful design includes the following: (a) the mathematical determination of the maximum cladding tube temperatures; (b) the determination of the maximum permissible fatigue strengths and creep strains of the materials; and (c) the safety distance between the nominal cladding tube hot spots and the permissible extreme cladding tube temperature. The maximum cladding tube temperatures occur on the top edge of the core and, due to radial power gradients, in the wrapper-wall region of a pin bundle. If grid spacers are now used for fixing the pins as in the SNR fuel elements, a careful check must be made of whether and to what degree temperature peaks in the region of the supports have an influence on the cladding tube design. Initial experimental investigations on a sodium-cooled pin bundle model of the SNR-300 fuel element were carried out to throw light on these special problems. This is reported in the following together with the results so far obtained. (U.K.)

Development of fuel performance and thermal hydraulic technology

International Nuclear Information System (INIS)

Jung, Youn Ho; Song, K. N.; Kim, H. K. and others

2000-03-01

Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)

Spent nuclear fuel system dynamic stability under normal conditions of transportation

Energy Technology Data Exchange (ETDEWEB)

Jiang, Hao; Wang, Jy-An John, E-mail: wangja@ornl.gov

2016-12-15

Highlights: • A conformational potential effect of fuel assembly contact interaction induced transient shock. • Complex vibration modes and vibration load intensity were observed from fuel assembly system. • The project was able to link the periodic transient shock to spent fuel fatigue strength reduction. - Abstract: In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside the cask during NCT. Dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly. To further evaluate the intensity of contact interaction induced by the local contacts’ impact loading at the spacer grid, detailed models of the actual spring and dimples of the spacer grids were created. The impacts between the fuel rod and springs and dimples were simulated with a 20 g transient shock load. The associated contact interaction intensities, in terms of reaction forces, were estimated from the finite element analyses (FEA) results. The bending moment estimated from the resultant stress on the clad under 20 g transient shock can be used to define the loading in cyclic integrated reversible-bending fatigue tester (CIRFT) vibration testing for the equivalent condition. To estimate the damage potential of the transient shock to the SNF vibration

The effect of vortex mixing grids on the behaviour of steam phase in fuel assemblies

International Nuclear Information System (INIS)

Sergeev, V.V.; Fedotovsky, V.S.; Shcherbakov, S. I.

2013-01-01

The paper examines the behavior of the steam phase in two-phase flow in the space between fuel elements of a transparent model of WWER FA with “Vortex”-type mixing grids. The model was a real subchannel scaled-up in the ratio of 5 to 1 and allowed for gas feeding through the walls of displacers to simulate boiling on fuel surface. Hydraulic resistances of intensifier simulators were determined during experiments and bubble paths were photographed at different velocities of the water flow. Experimental results made it possible to verify calculation models developed at SSC RF - IPPE for fuel assemblies with “Vortex”-type mixing grids. These models let calculate the effect of steam removal from fuel surface. (authors)

Diverse assemblies of the (4,4) grid layers exemplified in Zn(II)/Co(II) coordination polymers with dual linear ligands

International Nuclear Information System (INIS)

Liu, Guang-Zhen; Li, Xiao-Dong; Xin, Ling-Yun; Li, Xiao-Ling; Wang, Li-Ya

2013-01-01

Diverse (4,4) grid layers are exemplified in five two-dimensional coordination polymers with dual µ 2 -bridged ligands, namely, ([Zn(cbaa)(bpp)]·H 2 O) n (1), [Zn 2 (cbaa) 2 (bpy)] n (2), [Co 2 (cbaa) 2 (bpp) 2 ] n (3), [Co(cbaa)(bpp)] n (4), and [Co(bdaa)(bpp)(H 2 O) 2 ] n (5) (H 2 cbaa=4-carboxybenzeneacetic acid, bpp=1,3-di(4-pyridyl)propane, bpy=4,4′-bipyridyl, and H 2 bdaa=1,4-benzenediacrylic acid). For 1, two (4,4) grid layers with [ZnN 2 O 2 ] tetrahedron as the node are held together by lattice water forming a H-bonding bilayer. Individual (4,4) grid layer in 2 is based on (Zn 2 (OCO) 4 ) paddlewheel unit as the node. Two (4,4) grid layers with (Co 2 O(OCO) 2 ) dimer as the node are covalently interconnected by organic ligands affording a thick bilayer of 3 with new framework topology. The different entanglements between two coincident (4,4) grid layers with [CoN 2 O 4 ] octahedron as the node leads to two 2D→2D interpenetrated structures for 4 and 5. Furthermore, fluorescent properties of 1 and 2 as well as magnetic properties of 3 are investigated. - Graphical abstract: Diverse assemblies of the (4,4) grid layers with different network nodes forms five coordination polymers that are well characterized by IR, TGA, element analysis, fluorescent and magnetic measurement. - Highlights: • Diverse assemblies of the (4,4) grid layers with different structural units as the nodes. • A new topology type with the uninodal 6-connected net of (4 12 .5 2 .6) is found. • Intense fluorescence emissions with a rare blue-shift of 55 nm compared to free carboxylate ligand

Development of Geometry Optimization Methodology with In-house CFD code, and Challenge in Applying to Fuel Assembly

International Nuclear Information System (INIS)

Jeong, J. H.; Lee, K. L.

2016-01-01

The wire spacer has important roles to avoid collisions between adjacent rods, to mitigate a vortex induced vibration, and to enhance convective heat transfer by wire spacer induced secondary flow. Many experimental and numerical works has been conducted to understand the thermal-hydraulics of the wire-wrapped fuel bundles. There has been enormous growth in computing capability. Recently, a huge increase of computer power allows to three-dimensional simulation of thermal-hydraulics of wire-wrapped fuel bundles. In this study, the geometry optimization methodology with RANS based in-house CFD (Computational Fluid Dynamics) code has been successfully developed in air condition. In order to apply the developed methodology to fuel assembly, GGI (General Grid Interface) function is developed for in-house CFD code. Furthermore, three-dimensional flow fields calculated with in-house CFD code are compared with those calculated with general purpose commercial CFD solver, CFX. The geometry optimization methodology with RANS based in-house CFD code has been successfully developed in air condition. In order to apply the developed methodology to fuel assembly, GGI function is developed for in-house CFD code as same as CFX. Even though both analyses are conducted with same computational meshes, numerical error due to GGI function locally occurred in only CFX solver around rod surface and boundary region between inner fluid region and outer fluid region.

Apparatus for loading fuel rods into grids of nuclear fuel assemblies

International Nuclear Information System (INIS)

Shallenberger, J.M.; Ferlan, S.J.

1989-01-01

For use with a nuclear fuel assembly including support grids having cells for receiving fuel rods and with detents disposed within the respective cells for resiliently engaging and laterally supporting the fuel rods received therein, an apparatus is described for facilitating scratchless insertion of each fuel rod into cells of the support rids. The apparatus consists of: a thin-walled metallic tubular member which is long enough to extend through at least a majority of support grids, and is positionable so as to have its thin wall interposed, during insertion of each fuel rod, between the latter and the detents within the cells receiving it, the thin-walled tubular member having a substantially uniform wall thickness of not more than about 0.008 inch, an as-formed inner diameter substantially equal to the outer diameter of the fuel rod, and a longitudinal slit formed in the wall of the tubular member so as to render the wall resiliently deflectable in a diameter-reducing sense, the longitudinal slit having a width sufficient to preclude overlapping of the edges of the wall along the slit, and insufficient for any of the detents to enter the slit when the wall of the tubular member is in position between the detents and the fuel rod

Development of TVSA VVER-1000 fuel

International Nuclear Information System (INIS)

Samoilov, O.; Kaydalov, V.; Romanov, A.; Falkov, A.; Morozkin, O.; Sholin, E.

2013-01-01

The TVSA fuel assemblies with a rigid angle-piece skeleton operate at 21 VVER-1000 units of Kalinin NPP, and Ukrainian, and Czech and Bulgarian NPPs. The total of more than 6,000 TVSA fuel assemblies have been fabricated. High lifetime performance has been achieved, namely, the maximum FA burnup is 65 MW∙day/kgU; maximum fuel rod burnup is 72 MW∙day/kgU; the lifetime is 50,000 EFPH. The TVSA fuel assembly is being improved to enhance its technical and economic performance and competitiveness of the Russian fuel for the VVER-1000 reactor: 1) Reliability and safety are being enhanced; repairability is being ensured. 2) High burnup levels in fuel are being ensured. 3) The uranium content in FAs is being increased. 4) The operational life is being extended. 5) Thermal-technical characteristics of FAs are being improved. The basic TVSA fuel assembly design evolved into the TVSA-PLUS with the fuel column elongated by 150 mm. The TVSA-PLUS fuel assembly has been in operation since 2010 at Kalinin NPP power units; an eighteen-month cycle is implemented at the uprated power of 104%. The TVSA-12PLUS fuel assembly has been developed with an elongated fuel column, optimized spacer grid positions (the spacer grid pitch is 340 mm) and with ensuring higher rigidity for the skeleton. It is provided for that fuel rods with the elevated uranium content and mixing intensifier grids will be used. The TVSA-T is developed for VVER-1000 reactor cores at the Temelin NPP. The TVSA-T is characterized by a load-carrying skeleton formed with angle-pieces and combined spacer grids that incorporate mixer grids. The TVSA-T design won the international tender to supply fuel to the Temelin NPP in the Czech Republic, and currently Temelin NPP Unit 1 and 2 are operating with the cores fully loaded with TVSA-Ts

Flow behavior of droplets downstream of the spacer

International Nuclear Information System (INIS)

Kodama, Eiichiro; Morishita, Kiyohide; Aritomi, Masanori; Yano, Takashi

1998-01-01

The fuel spacer, of which role is to maintain an appropriate rod-to-rod clearance, is one of the components of a Boiling Water Reactor (BWR) fuel rod bundles. The fuel spacer influences flow characteristics of the liquid film in fuel rod bundles, so that its geometry influences greatly thermal hydraulics such as critical power and pressure drop therein. The purpose of this study is to clarify the effect of the spacer geometry on the core flow split downstream of the spacer. Phase Doppler Anemometry (PDA) was used for their meausrement under the conditions of a small amount of droplets in mist flows. From the experimental results, the normalized droplet velocity profiles with a spacer were split by the spacer and were different between a wider and a narrower regions in the channel, however, they became uniform at the distance far 100mm from the spacer. In the case without a spacer, the velocity was monotonously increasing nearer the rod surface with going toward the center of the channel. In the case with a spacer, the velocity profile downstream of the spacer changed in the narrower region of the channel. This tendency became more remarkable with thickening the spacer and widening clearance between the spacer and the wall. In this paper, 'drift' velocity effect was applied for the spacer model, due to the gas flows were split by the spacer which is based on the momentum balance between the narrower and wider channels. This model was confirmed from the experimental results that the droplet flowed from a wider region to a narrower one. This drift effect appeared more strongly as the spacer became thicker and the clearance did narrower. The analytical results explained qualitatively the measured ones. It is clarified that the drift effect proposed in this work was a dominant factor on droplet deposition downstream of the spacer

Core baffle for nuclear reactors

International Nuclear Information System (INIS)

Machado, O.J.; Berringer, R.T.

1977-01-01

The invention concerns the design of the core of a LWR with a large number of fuel assemblies formed by fuel rods and kept in position by spacer grids. According to the invention, at the level of the spacer grids match plates are mounted with openings so the flow of coolant directed upwards will not be obstructed and a parallel bypass will be obtained in the space between the core barrel and the baffle plates. In case of an accident, this configuration reduces or avoids damage from overpressure reactions. (HP) [de

Fouling Resilient Perforated Feed Spacers for Membrane Filtration

KAUST Repository

Kerdi, Sarah

2018-04-24

The improvement of feed spacers with optimal geometry remains a key challenge for spiral-wound membrane systems in water treatment due to their impact on the hydrodynamic performance and fouling development. In this work, novel spacer designs are proposed by intrinsically modifying cylindrical filaments through perforations. Three symmetric perforated spacers (1-Hole, 2-Hole, and 3-Hole) were in-house 3D-printed and experimentally evaluated in terms of permeate flux, feed channel pressure drop and membrane fouling. Spacer performance is characterized and compared with standard no perforated (0-Hole) design under constant feed pressure and constant feed flow rate. Perforations in the spacer filaments resulted in significantly lowering the net pressure drop across the spacer filled channel. The 3-Hole spacer was found to have the lowest pressure drop (50% - 61%) compared to 0-Hole spacer for various average flow velocities. Regarding permeate flux production, the 0-Hole spacer produced 5.7 L.m-2.h-1 and 6.6 L.m-2.h-1 steady state flux for constant pressure and constant feed flow rate, respectively. The 1-Hole spacer was found to be the most efficient among the perforated spacers with 75% and 23% increase in permeate production at constant pressure and constant feed flow, respectively. Furthermore, membrane surface of 1-Hole spacer was found to be cleanest in terms of fouling, contributing to maintain higher permeate flux production. Hydrodynamic understanding of these perforated spacers is also quantified by performing Direct Numerical Simulation (DNS). The performance enhancement of these perforated spacers is attributed to the formation of micro-jets in the spacer cell that aided in producing enough unsteadiness/turbulence to clean the membrane surface and mitigate fouling phenomena. In the case of 1-Hole spacer, the unsteadiness intensity at the outlet of micro-jets and the shear stress fluctuations created inside the cells are higher than those observed with

Fouling resilient perforated feed spacers for membrane filtration.

Science.gov (United States)

Kerdi, Sarah; Qamar, Adnan; Vrouwenvelder, Johannes S; Ghaffour, Noreddine

2018-04-24

The improvement of feed spacers with optimal geometry remains a key challenge for spiral-wound membrane systems in water treatment due to their impact on the hydrodynamic performance and fouling development. In this work, novel spacer designs are proposed by intrinsically modifying cylindrical filaments through perforations. Three symmetric perforated spacers (1-Hole, 2-Hole, and 3-Hole) were in-house 3D-printed and experimentally evaluated in terms of permeate flux, feed channel pressure drop and membrane fouling. Spacer performance is characterized and compared with standard no perforated (0-Hole) design under constant feed pressure and constant feed flow rate. Perforations in the spacer filaments resulted in significantly lowering the net pressure drop across the spacer filled channel. The 3-Hole spacer was found to have the lowest pressure drop (50%-61%) compared to 0-Hole spacer for various average flow velocities. Regarding permeate flux production, the 0-Hole spacer produced 5.7 L m -2 .h -1 and 6.6 L m -2 .h -1 steady state flux for constant pressure and constant feed flow rate, respectively. The 1-Hole spacer was found to be the most efficient among the perforated spacers with 75% and 23% increase in permeate production at constant pressure and constant feed flow, respectively. Furthermore, membrane surface of 1-Hole spacer was found to be cleanest in terms of fouling, contributing to maintain higher permeate flux production. Hydrodynamic understanding of these perforated spacers is also quantified by performing Direct Numerical Simulation (DNS). The performance enhancement of these perforated spacers is attributed to the formation of micro-jets in the spacer cell that aided in producing enough unsteadiness/turbulence to clean the membrane surface and mitigate fouling phenomena. In the case of 1-Hole spacer, the unsteadiness intensity at the outlet of micro-jets and the shear stress fluctuations created inside the cells are higher than those

Enhanced Westinghouse WWER-1000 fuel design for Ukraine reactors

International Nuclear Information System (INIS)

Dye, M.; Shah, H.

2015-01-01

Westinghouse has completed design, development, and region quantity delivery of an enhanced Westinghouse fuel assembly for WWER-1000 reactors to support continued safe reactor operations. The enhanced design builds on the successful performance of an earlier generation design which has operated in the South Ukraine 3 reactor for multiple cycles without any fuel rod failures. Incorporated design enhancements include a thicker spacer grid outer strap, an enhanced spacer grid outer strap profile to limit the risk for, and impact of, mechanical interaction/interference with coresident fuel, an all Alloy 718 grid structure for improved stability and strength, and improvements to the top and bottom nozzles. Capable of meeting increased lateral loads generated from using a higher axial trip limit for the refueling machine crane, the design was verified by extensive mechanical and thermalhydraulic testing, which included a newly developed fuel assembly-to-fuel assembly handling test rig to assess performance during bounding core loading and unloading conditions. Through these extensive design enhancements and comprehensive testing program, the enhanced WWER-1000 design provides additional performance, handling, and reliability margins for safe reactor operation. (authors)

Most advanced HTP fuel assembly design for EPR

International Nuclear Information System (INIS)

Francillon, Eric; Kiehlmann, Horst-Dieter

2006-01-01

End 2003, the Finnish electricity utility Teollisuuden Voima Oy (TVO) signed the contract for building an EPR in Olkiluoto (Finland). Mid 2004, the French electricity utility EDF selected an EPR to be built in France. In 2005, Framatome ANP, an AREVA and Siemens company, announced that they will be pursuing a design certification in the U.S. The EPR development is based on the latest PWR product lines of former Framatome (N4) and Siemens Nuklear (Konvoi). As an introductory part, different aspects of the EPR core characteristics connected to fuel assembly design are presented. It includes means of ensuring reactivity control like hybrid AIC/B4C control rod absorbers and gadolinium as burnable absorber integrated in fuel rods, and specific options for in-core instrumentation, such as Aeroball type instrumentation. Then the design requirements for the EPR fuel assembly are presented in term of very high burnup capacity, rod cladding and fuel assembly reliability. Framatome ANP fuel assembly product characteristics meeting these requirements are then described. EPR fuel assembly design characteristics benefit from the experience feedback of the latest fuel assembly products designed within Framatome ANP, leading to resistance to assembly deformation, high fuel rod restraint and prevention of handling hazards. EPR fuel assembly design features the best components composing the cornerstones of the upgraded family of fuel assemblies that FRAMATOME ANP proposes today. This family is based on a set of common characteristics and associated features, which include the HMP grid as bottom end spacer, the MONOBLOC guide tube and the Robust FUELGUARD as lower tie plate, the use of the M5 Alloy, as cladding and structure material. This fully re-crystallized, ternary Zr-Nb-O alloy produces radically improved in-reactor corrosion, very low hydrogen uptake and growth and an excellent creep behavior, which are described there. EPR fuel assembly description also includes fuel rod

Diverse assemblies of the (4,4) grid layers exemplified in Zn(II)/Co(II) coordination polymers with dual linear ligands

Energy Technology Data Exchange (ETDEWEB)

Liu, Guang-Zhen; Li, Xiao-Dong; Xin, Ling-Yun; Li, Xiao-Ling [College of Chemistry and Chemical Engineering, Luoyang Normal University, Luoyang, Henan 471022 (China); Wang, Li-Ya, E-mail: wlya@lynu.edu.cn [College of Chemistry and Chemical Engineering, Luoyang Normal University, Luoyang, Henan 471022 (China); College of Chemistry and Pharmacy Engineering, Nanyang Normal University, Nanyang, Henan 473061 (China)

2013-07-15

Diverse (4,4) grid layers are exemplified in five two-dimensional coordination polymers with dual µ{sub 2}-bridged ligands, namely, ([Zn(cbaa)(bpp)]·H{sub 2}O){sub n} (1), [Zn{sub 2}(cbaa){sub 2}(bpy)]{sub n} (2), [Co{sub 2}(cbaa){sub 2}(bpp){sub 2}]{sub n} (3), [Co(cbaa)(bpp)]{sub n} (4), and [Co(bdaa)(bpp)(H{sub 2}O){sub 2}]{sub n} (5) (H{sub 2}cbaa=4-carboxybenzeneacetic acid, bpp=1,3-di(4-pyridyl)propane, bpy=4,4′-bipyridyl, and H{sub 2}bdaa=1,4-benzenediacrylic acid). For 1, two (4,4) grid layers with [ZnN{sub 2}O{sub 2}] tetrahedron as the node are held together by lattice water forming a H-bonding bilayer. Individual (4,4) grid layer in 2 is based on (Zn{sub 2}(OCO){sub 4}) paddlewheel unit as the node. Two (4,4) grid layers with (Co{sub 2}O(OCO){sub 2}) dimer as the node are covalently interconnected by organic ligands affording a thick bilayer of 3 with new framework topology. The different entanglements between two coincident (4,4) grid layers with [CoN{sub 2}O{sub 4}] octahedron as the node leads to two 2D→2D interpenetrated structures for 4 and 5. Furthermore, fluorescent properties of 1 and 2 as well as magnetic properties of 3 are investigated. - Graphical abstract: Diverse assemblies of the (4,4) grid layers with different network nodes forms five coordination polymers that are well characterized by IR, TGA, element analysis, fluorescent and magnetic measurement. - Highlights: • Diverse assemblies of the (4,4) grid layers with different structural units as the nodes. • A new topology type with the uninodal 6-connected net of (4{sup 12}.5{sup 2}.6) is found. • Intense fluorescence emissions with a rare blue-shift of 55 nm compared to free carboxylate ligand.

EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

Directory of Open Access Journals (Sweden)

S. M. Dmitriev

2016-01-01

Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

Spacer for supporting fuel element boxes

International Nuclear Information System (INIS)

Wild, E.

1979-01-01

A spacer plate unit arranged externally on each side and at a predetermined level of a polygonal fuel element box for mutually supporting, with respect to one another, a plurality of the fuel element boxes forming a fuel element bundle, is formed of a first and a second spacer plate part each having the same length and the same width and being constituted of unlike first and second materials, respectively. The first and second spacer plate parts of the several spacer plate units situated at the predetermined level are arranged in an alternating continuous series when viewed in the peripheral direction of the fuel element box, so that any two spacer plate units belonging to face-to-face oriented sides of two adjoining fuel element boxes in the fuel element bundle define interfaces of unlike materials

Fuel rod-to-support contact pressure and stress measurement for CHASNUPP-1(PWR) fuel

International Nuclear Information System (INIS)

Waseem; Elahi, N.; Siddiqui, A.; Murtaza, G.

2011-01-01

Research highlights: → A detailed finite element model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. → The spring hold-down force is calculated using the contact pressure obtained from the FE model. → Experiment has also been conducted in the same environment for the measurement of this force. → The spring hold-down force values obtained from both studies confirm the validation of this analysis. → The stress obtained through this analysis is less than the yield strength of spacer grid material, thus fulfils the structural integrity criteria of grid. - Abstract: This analysis has been made in an attempt to measure the contact pressure of the PWR fuel assembly spacer grid spring and to verify its structural integrity at room temperature in air. A detailed finite element (FE) model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The FE model of a fuel rod-to-support system is produced with shell and contact elements. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies are compared, which show good agreement, and in turn confirm the validation of this analysis. The Stress obtained through this analysis is less than the yield strength of spacer grid material (Inconel-718), thus fulfils the structural integrity criteria of grid.

Fuel rod-to-support contact pressure and stress measurement for CHASNUPP-1(PWR) fuel

Energy Technology Data Exchange (ETDEWEB)

Waseem, E-mail: wazim_me@hotmail.co [Directorate General Nuclear Power Fuel, Pakistan Atomic Energy Commission, P.O. Box No. 1847, Islamabad 44000 (Pakistan); Elahi, N.; Siddiqui, A.; Murtaza, G. [Directorate General Nuclear Power Fuel, Pakistan Atomic Energy Commission, P.O. Box No. 1847, Islamabad 44000 (Pakistan)

2011-01-15

Research highlights: A detailed finite element model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies confirm the validation of this analysis. The stress obtained through this analysis is less than the yield strength of spacer grid material, thus fulfils the structural integrity criteria of grid. - Abstract: This analysis has been made in an attempt to measure the contact pressure of the PWR fuel assembly spacer grid spring and to verify its structural integrity at room temperature in air. A detailed finite element (FE) model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The FE model of a fuel rod-to-support system is produced with shell and contact elements. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies are compared, which show good agreement, and in turn confirm the validation of this analysis. The Stress obtained through this analysis is less than the yield strength of spacer grid material (Inconel-718), thus fulfils the structural integrity criteria of grid.

Intermediate flow mixing nonsupport grid for BWR fuel assembly

International Nuclear Information System (INIS)

Taleyarkhan, R.P.

1987-01-01

An intermediate flow mixing nonsupport grid is described for use in a nuclear reactor fuel assembly containing an array of elongated fuel rods. The grid comprises: (a) interleaved inner straps arranged in an egg-crate configuration to define inner cell openings for receiving respective ones of the fuel rods. The inner straps have outer terminal end portions; (b) an outer peripheral strap attached to the respective terminal end portions of the inner straps to define perimeter cell openings for receiving other ones of the fuel rods. The inner straps and outer strap together have opposite upstream and downstream sides; (c) a first group of mixing vanes disposed at the downstream side and being attached on portions of the outer strap and on respective portions of the inner straps. Together with the outer strap portions, they define the perimeter cell openings. Each of the mixing vanes of the first group extend generally in a downstream direction and inwardly toward the perimeter cell openings for deflecting coolant flowing; and (d) a second group of mixing vanes disposed at the downstream side and being attached on other portions of the inner straps. Together with the respective portions, they define the inner cell openings. Each of the mixing vanes of the second group extend generally in a downstream direction and inwardly toward the inner cell openings for deflecting coolant flowing therethrough; (e) the mixing vanes of the second group are substantially smaller in size than the mixing vanes of the first group so as to generate substantially less turbulence in the portions of the coolant flowing through the inner cell openings than in the portions of the coolant flowing through the perimeter cell openings

Rehme correlation for spacer pressure drop compared to XT-ADS rod bundle simulations and water experiment

International Nuclear Information System (INIS)

Batta, A.; Class, A.; Litfin, K.; Wetzel, T.

2011-01-01

The Rehme correlation is the most common formula to estimate the pressure drop of spacers in the design phase of new bundle geometries. It is based on considerations of momentum losses and takes into account the obstruction of the flow cross section but it ignores the geometric details of the spacer design. Within the framework of accelerator driven sub-critical reactor systems (ADS), heavy-liquid-metal (HLM) cooled fuel assemblies are considered. At the KArlsruhe Liquid metal LAboratory (KALLA) of the Karlsruhe Institute of Technology a series of experiments to quantify both pressure losses and heat transfer in HLM-cooled rod bundles are performed. The present study compares simulation results obtained with the commercial CFD code Star-CCM to experiments and the Rehme correlation. It can be shown that the Rehme correlation, simulations and experiments all yield similar trends, but quantitative predictions can only be delivered by the CFD which takes into account the full geometric details of the spacer geometry. (orig.)

The effect of grid assembly mixing vanes on critical heat flux values and azimuthal location in fuel assemblies

International Nuclear Information System (INIS)

De Crecy, F.

1994-01-01

Critical heat flux (CHF) is one of the limiting phenomena for a PWR. It has been widely studied for years, but many facts are still not satisfactorily understood. This paper deals with the effect of the grid assembly mixing vanes on both the value of the CHF and the azimuthal location of the departure from nucleate boiling (DNB). A series of experimental studies was performed on electrically heated, 5x5 square pitched, vertical rod bundles. Two specific grid assembly designs were used: with and without mixing vanes. DNB was detected by eight thermocouples welded internally in each rod at the same level in order to determine the azimuthal location. The coolant was Freon-12 flowing upwards to simulate high pressure water (as defined by Stevens). Single-phase flow experiments were also conducted to measure the exit temperature field in order to obtain the mixing coefficients for subchannel analysis.The results show very clearly that the mixing vanes have a significant effect on both the DNB azimuthal location and the CHF value. - Without mixing vanes, DNB occurs mainly on the most central rod and preferentially at the azimuthal location facing the adjacent rod. - With mixing vanes, DNB can occur on any of the nine central rods and is distributed in an apparently random way around the rod. -The effect of the mixing vanes on CHF is dramatic and depends a great deal on the parameter range (pressure, local mass velocity and local quality). Generally speaking, CHF with mixing vanes is significantly higher than without mixing vanes, but this effect can be inverted in some cases.In order to understand this fact more clearly, it is necessary to perform detailed analysis of subchannel behavior. Indeed, the analyses show that the magnitude of this effect is closely related to the mixing coefficients used. These mixing coefficients, estimated from the single-phase flow experiments, are subject to large uncertainties in two-phase flow. ((orig.))

Fouling Resilient Perforated Feed Spacers for Membrane Filtration

KAUST Repository

Kerdi, Sarah; Qamar, Adnan; Vrouwenvelder, Johannes S.; Ghaffour, NorEddine

2018-01-01

drop across the spacer filled channel. The 3-Hole spacer was found to have the lowest pressure drop (50% - 61%) compared to 0-Hole spacer for various average flow velocities. Regarding permeate flux production, the 0-Hole spacer produced 5.7 L.m-2.h-1

Neutron detector assembly

International Nuclear Information System (INIS)

Hanai, Koi; Shirayama, Shinpei.

1978-01-01

Purpose: To prevent gamma-ray from leaking externally passing through the inside of a neutron detector assembly. Constitution: In a neutron detector assembly having a protection pipe formed with an enlarged diameter portion which serves also as a spacer, partition plates with predetermined width are disposed at the upper and the lower portions in this expanded portion. A lot of metal particles are filled into spaces formed by the partition plates. In such a structure, the metal particles well-absorb the gamma-rays from above and convert them into heat to provide shielding for the gamma-rays. (Horiuchi, T.)

Antibiotic impregnated total femur spacers: a technical tip

Directory of Open Access Journals (Sweden)

Colin D. Canham, MD

2018-03-01

Full Text Available Simultaneous prosthetic joint infection of ipsilateral hip and knee arthroplasties is often accompanied by significant bone loss and presents a challenging reconstructive problem. Two-stage reconstruction is favored and requires the placement of a total femur spacer, which is not a commercially available device. We describe a surgical technique, reporting on 2 cases in which a customized total femur antibiotic impregnated spacer was created by combining an articulating knee spacer and an articulating hip spacer with a reinforced cement dowel construct connecting the 2 spacers. Custom total femoral spacers are useful in the management of infected femoral megaprostheses and cases with ipsilateral injected hip and knee arthroplasties and severe femoral bone loss. Keywords: total femur spacer, revision arthroplasty, total hip arthroplasty, total knee arthroplasty, prosthetic joint infection

Fuel assembly

International Nuclear Information System (INIS)

Nakatsuka, Masafumi; Matsuzuka, Ryuji.

1976-01-01

Object: To provide a fuel assembly which can decrease pressure loss of coolant to uniform temperature. Structure: A sectional area of a flow passage in the vicinity of an inner peripheral surface of a wrapper tube is limited over the entire length to prevent the temperature of a fuel element in the outermost peripheral portion from being excessively decreased to thereby flatten temperature distribution. To this end, a plurality of pincture-frame-like sheet metals constituting a spacer for supporting a fuel assembly, which has a plurality of fuel elements planted lengthwise and in given spaced relation within the wrapper tube, is disposed in longitudinal grooves and in stacked fashion to form a substantially honeycomb-like space in cross section. The fuel elements are inserted and supported in the space to form a fuel assembly. (Kamimura, M.)

On numerical simulation of fuel assembly bow in pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Horváth, Ákos, E-mail: akoshorvath@t-online.hu [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany); Budapest University of Technology and Economics, Department of Aircraft and Ships, Stoczek Street 6, Building J, H-1111 Budapest (Hungary); Dressel, Bernd [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany)

2013-12-15

Highlights: • Simulation of fuel assembly bow by coupled CFD and finite element method. • Comparison of calculated and experimentally measured bow shapes. • Investigation of boundary condition effect on bow pattern of a fuel assembly row. • Highlighting importance of consideration of fluid–structure interaction. • Assessment of flow redistribution within the fuel assembly row model. - Abstract: Fuel assembly bow in pressurized water reactor cores is largely triggered by lateral hydraulic forces together with creep processes generated by neutron flux. A detailed understanding of the flow induced bow behaviour is, therefore, an important issue. The experimental feedbacks and laboratory tests on fuel assembly bow show that it is characterized to a high degree by fluid–structure interaction (FSI) effects, therefore, consideration of FSI is essential and indispensable in full comprehension of the bow mechanism. In the present study, coupled computational fluid dynamics (CFD) and finite element simulations are introduced, calculating fuel assembly deformation under different conditions as a quasi-stationary phenomenon. The aim has been, on the one hand, to develop such a simplified fuel assembly CFD model, which allows set up of fuel assembly rows without loosing its main hydraulic characteristic; on the other hand, to investigate the bow pattern of a given fuel assembly row under different boundary conditions. The former one has been achieved by comparing bow shapes obtained with different fuel assembly (spacer grid) modelling approaches and mesh resolutions with experimental data. In the second part of the paper a row model containing 7.5 fuel assemblies is introduced, investigating the effect of flow distribution at inlet and outlet boundary regions on fuel assembly bow behaviour. The post processing has been focused on the bow pattern, lateral hydraulic forces, and horizontal flow distribution. The results have revealed importance of consideration of

Casette for storage of spent fuel assemblies

International Nuclear Information System (INIS)

Ericsson, S.

1992-01-01

Describes a design of a casette for spent fuel storage in a fuelstorage pool. The new design, based on flexible spacers, allows the fuel assemblies to be packed more compact and the fuel storage pool used in a more economic way

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)

Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

Science.gov (United States)

Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

2017-07-01

Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes).

Nuclear reactor fuel element assemblies

International Nuclear Information System (INIS)

Raven, L.F.

1975-01-01

A spacer grid for a nuclear fuel element comprises a plurality of cojointed cylindrical ferrules adapted to receive a nuclear fuel pin. Each ferrule has a pair of circumferentially spaced rigid stop members extending inside the ferrule and a spring locating member attached to the ferrule and also extending from the ferrule wall inwardly thereof at such a circumferential spacing relative to the rigid stop members that the line of action of the spring locating member passes in opposition to and between the rigid stop members which lie in the same diametric plane. At least some of the cylindrical ferrules have one rim shaped to promote turbulence in fluid flowing through the grid. (Official Gazette)

Fabrication of a pressurized water reactor fuel element prototype with Zy-control rod guide tubes

International Nuclear Information System (INIS)

Bezold, H.; Romeiser, H.J.

1978-10-01

A prototype fuel assembly with zircaloy guide was fabricated by mass production methods. The fastening of the Inconel spacer grids to the guide tubes and the transition joint for fixing the tubes to the stainless stell upper end-fitting of the assembly were investigated. Tools and welding devices were developed for the construction of the skeleton. (orig.) [de

Detection and replacement of crud accumulating fuel assemblies at the Loviisa-2 reactor

International Nuclear Information System (INIS)

Antila, M.

1995-01-01

At unit 2 of Loviisa NPP fuel bundle outlet temperatures had been increasing slowly from the beginning of November 1994 mainly at the six lead fuel bundles with new zirconium spacer grids (Z). The unit was taken to cold shut-down and started up again in January 1995. After a few days the increase of the outlet temperatures was confirmed to continue. It was decided to shut down the reactor again to examine and remove the new Z-bundles. One bundle was taken to closer investigations in the poolside inspection stand. The hexagonal shroud tube was removed and quite a lot of crud was detected on the lower surfaces of the spacer grids. All the six Z-bundles were removed and replaced by older bundles with steel spacers. A small reshuffling was performed to control the radial power distribution. The cause for the partial flow blockage causing the increasing temperatures was accumulation of crud particularly on the Zr-spacers. The root cause of the accumulation of crud has not yet been definitely cleared. It seems evident that it is related to the decontamination of the primary circuit performed last August. The extensive incore temperature measurement instrumentation and reactor core monitoring system had a central role in the early detection of the anomaly. Thus core thermal margins were not violated and possible fuel failures could be avoided. (orig.) (3 refs., 5 figs.)

Comparative studies of 111In-labeled monoclonal antibody using spacer-containing and non-spacer bifunctional chelates. 2

International Nuclear Information System (INIS)

Sun, Baofu

1994-01-01

Indium-111-labeled A7 monoclonal antibodies using two spacer-containing chelates, succinimido-EGS-DTPA (EGS-DTPA: diester spacer) and maleimido-C10-Bz-EDTA (C10-Bz-EDTA: hydrocarbon spacer) were investigated in human LS180 colon tumor bearing nude mice and were compared with two non-spacer chelates, cyclic DTPA dianhydride (cDTPAA) and isothiocyanatobenzyl-EDTA (SCN-Bz-EDTA). Compared with immunoconjugates using non-spacer chelates, immunoconjugates using spacer-containing chelates, especially C10-Bz-EDTA-A7 showed lower 111 In activity in normal organs. The radioactivity in the liver for C10-Bz-EDTA-A7 decreased continuously till 96 hrs postinjection, however, this liver radioactivity for EGS-DTPA-A7 showed little change after 24 hrs. Moreover, in liver subcellular distribution study, EGS-DTPA-A7 showed a higher activity retention in mitochondrial fraction which contained lysosome, a place for metabolizing and storing of 111 In labeled antibody, than that of C10-Bz-EDTA-A7. The C10-Bz-EDTA-A7 conjugate demonstrated more preferable tumor-to-non tumor contrast on the scintigrams than that found with other three immunoconjugates. Up to 96 hrs postinjection, tumor bearing nude mice injecting with immunoconjugates using spacer-containing chelates exreted twice radioactivity from whole body than that excreted by using non-spacer chelates. Interestingly, different from other three chelates, C10-Bz-EDTA-A7 were mainly excreted via feces. We conclude that the decrease of radioactivity in normal tissues in the case of EGS-DTPA-A7 was due to the rapid decrease of activity in the blood, while in the case of C10-Bz-EDTA-A7 it was due to the quickly excreted small metabolite through faces. 111 In labeled C10-Bz-EDTA conjugate is superior, at least when conjugated with A7, to other three chelate conjugates used in this study. (author)

Fuel assembly loads during a hypothetical blowdown event in a PWR

International Nuclear Information System (INIS)

Stabel, J.; Bosanyi, B.; Kim, J.D.

1991-01-01

As a consequence of a hypothetical sudden break of the main coolant pipe of a PWR, RPV-internals and fuel assemblies (FA's) are undergoing horizontal and vertical motions. FA's may impact against each other, against core shroud or against lower core support. The corresponding impact loads must be absorbed by the FA spacer grids and guide thimbles. In this paper FA-loads are calculated with and without consideration of Fluid-Structure-Interaction (FSI) effects for assumed different break sizes of the main coolant pipe. The analysis has been performed for a hypothetical cold leg break of a typical SIEMENS-4 loop plant. For this purpose the codes DAPSY/DAISY (GRS, Germany) were coupled with the structural code KWUSTOSS (SIEMENS). It is shown that the FA loads obtained in calculations with consideration of FSI effects are by a factor of 2-4 lower than those obtained in the corresponding calculations without consideration of FSI. (author)

Diversity in a Polymicrobial Community Revealed by Analysis of Viromes, Endolysins and CRISPR Spacers.

Directory of Open Access Journals (Sweden)

Michelle Davison

Full Text Available The polymicrobial biofilm communities in Mushroom and Octopus Spring in Yellowstone National Park (YNP are well characterized, yet little is known about the phage populations. Dominant species, Synechococcus sp. JA-2-3B'a(2-13, Synechococcus sp. JA-3-3Ab, Chloroflexus sp. Y-400-fl, and Roseiflexus sp. RS-1, contain multiple CRISPR-Cas arrays, suggesting complex interactions with phage predators. To analyze phage populations from Octopus Spring biofilms, we sequenced a viral enriched fraction. To assemble and analyze phage metagenomic data, we developed a custom module, VIRITAS, implemented within the MetAMOS framework. This module bins contigs into groups based on tetranucleotide frequencies and CRISPR spacer-protospacer matching and ORF calling. Using this pipeline we were able to assemble phage sequences into contigs and bin them into three clusters that corroborated with their potential host range. The virome contained 52,348 predicted ORFs; some were clearly phage-like; 9319 ORFs had a recognizable Pfam domain while the rest were hypothetical. Of the recognized domains with CRISPR spacer matches, was the phage endolysin used by lytic phage to disrupt cells. Analysis of the endolysins present in the thermophilic cyanophage contigs revealed a subset of characterized endolysins as well as a Glyco_hydro_108 (PF05838 domain not previously associated with sequenced cyanophages. A search for CRISPR spacer matches to all identified phage endolysins demonstrated that a majority of endolysin domains were targets. This strategy provides a general way to link host and phage as endolysins are known to be widely distributed in bacteriophage. Endolysins can also provide information about host cell wall composition and have the additional potential to be used as targets for novel therapeutics.

The potential of standard and modified feed spacers for biofouling control

KAUST Repository

Araú jo, Paula A.; Kruithof, Joop C.; van Loosdrecht, Mark C.M.; Vrouwenvelder, Johannes S.

2012-01-01

The impact of feed spacers on initial feed channel pressure (FCP) drop, FCP increase and biomass accumulation has been studied in membrane fouling simulators using feed spacers applied in commercially available nanofiltration and reverse osmosis spiral wound membrane modules. All spacers had a similar geometry.Our studies showed that biofouling was not prevented by (i) variation of spacer thickness, (ii) feed spacer orientation, (iii) feed spacer coating with silver, copper or gold and (iv) using a biostatic feed spacer. At constant feed flow, a lower FCP and FCP increase were observed for a thicker feed spacer. At constant linear flow velocity, roughly the same FCP development and biomass accumulation were found irrespective of the feed spacer thickness: hydrodynamics and substrate load were more important for development and impact of biofouling than the thickness of currently applied spacers. Use of biostatic and metal coated spacers were not effective for biofouling control. The same small reduction of biofouling rate was observed with copper and silver coated spacers as well as uncoated 45° rotated spacers.The studied modified spacers were not effective for biofouling prevention and control. The impact of biofouling on FCP increase was reduced significantly by a lower linear flow velocity, while spacer orientation and spacer thickness in membrane modules had a smaller but still significant effect. © 2012 Elsevier B.V.

The potential of standard and modified feed spacers for biofouling control

KAUST Repository

Araújo, Paula A.

2012-06-01

The impact of feed spacers on initial feed channel pressure (FCP) drop, FCP increase and biomass accumulation has been studied in membrane fouling simulators using feed spacers applied in commercially available nanofiltration and reverse osmosis spiral wound membrane modules. All spacers had a similar geometry.Our studies showed that biofouling was not prevented by (i) variation of spacer thickness, (ii) feed spacer orientation, (iii) feed spacer coating with silver, copper or gold and (iv) using a biostatic feed spacer. At constant feed flow, a lower FCP and FCP increase were observed for a thicker feed spacer. At constant linear flow velocity, roughly the same FCP development and biomass accumulation were found irrespective of the feed spacer thickness: hydrodynamics and substrate load were more important for development and impact of biofouling than the thickness of currently applied spacers. Use of biostatic and metal coated spacers were not effective for biofouling control. The same small reduction of biofouling rate was observed with copper and silver coated spacers as well as uncoated 45° rotated spacers.The studied modified spacers were not effective for biofouling prevention and control. The impact of biofouling on FCP increase was reduced significantly by a lower linear flow velocity, while spacer orientation and spacer thickness in membrane modules had a smaller but still significant effect. © 2012 Elsevier B.V.

Fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Gjertsen, R.K.; Tower, S.N.; Huckestein, E.A.

1982-01-01

A fuel assembly for a nuclear reactor comprises a 5x5 array of guide tubes in a generally 20x20 array of fuel elements, the guide tubes being arranged to accommodate either control rods or water displacer rods. The fuel assembly has top and bottom Inconel (Registered Trade Mark) grids and intermediate Zircaloy grids in engagement with the guide tubes and supporting the fuel elements and guide tubes while allowing flow of reactor coolant through the assembly. (author)

A non-electrostatic spacer for aerosol delivery

DEFF Research Database (Denmark)

Bisgaard, H; Anhøj, J; Klug, B

1995-01-01

to 6 years, suspected to have asthma the non-electrostatic spacer delivered a mean total dose of budesonide aerosol of 39% of the nominal dose, which was significantly higher than the Babyhaler (28%), the Nebuhaler (21%), and the AeroChamber (19%). These differences were most pronounced in children......A pear shaped non-electrostatic spacer, composed of steel with a volume of 250 ml and equipped with a facemask containing integrated inlet and outlet valves for inspiration and expiration, was compared with three plastic spacers. The plastic spacers were primed with repeated puffs from a budesonide...... was 27 seconds and independent of the use of p-MDI. In vitro the maximum dose of budesonide from a p-MDI, expressed as a percentage of the nominal dose, was 56% from the non-electrostatic spacer, 61% from the Nebuhaler, 45% from the Babyhaler, and 30% from the AeroChamber. In 124 children, age 6 months...

A non-electrostatic spacer for aerosol delivery

DEFF Research Database (Denmark)

Bisgaard, H; Anhøj, J; Klug, B

1995-01-01

A pear shaped non-electrostatic spacer, composed of steel with a volume of 250 ml and equipped with a facemask containing integrated inlet and outlet valves for inspiration and expiration, was compared with three plastic spacers. The plastic spacers were primed with repeated puffs from a budesonide...... pressurised metered dose inhaler (p-MDI) to minimise the electrostatic charge on the plastic. The procedure prolonged the half life (t1/2) of the aerosol in the Nebuhaler from nine to 32 seconds. A normal cleaning procedure reduced the aerosol t1/2 back to baseline. The t1/2 of the aerosol in the metal spacer...... was 27 seconds and independent of the use of p-MDI. In vitro the maximum dose of budesonide from a p-MDI, expressed as a percentage of the nominal dose, was 56% from the non-electrostatic spacer, 61% from the Nebuhaler, 45% from the Babyhaler, and 30% from the AeroChamber. In 124 children, age 6 months...

Positioning and locking device for fuel pin to grid attachment

International Nuclear Information System (INIS)

Frick, T.M.; Wineman, A.L.

1976-01-01

A positioning and locking device for fuel pin to grid attachment provides an inexpensive means of positively positioning and locking the individual fuel pins which make up the driver fuel assemblies used in nuclear reactors. The device can be adapted for use with a currently used attachment grid assembly design and insures that the pins remain in their proper position throughout the in-reactor life of the assembly. This device also simplifies fuel bundle assembly in that a complete row of fuel pins can be added to the bundle during each step of assembly. 8 claims, 8 drawing figures

Nuclear fuel assembly

International Nuclear Information System (INIS)

Ueda, Tomihiro.

1970-01-01

The present invention relates to fuel assemblies employing wire wrap spacers for retaining uniform spatial distribution between fuel elements. Clad fuel elements are helically wound in the oxial direction with a wave-formed wire strand. The strand is therefore provided with spring action which permits the fuel elements to expand freely in the axial and radial directions so as to retain proper spacing and reduce stresses due to thermal deformation. (Ownes, K.J.)

Assembling the Streptococcus thermophilus clustered regularly interspaced short palindromic repeats (CRISPR) array for multiplex DNA targeting.

Science.gov (United States)

Guo, Lijun; Xu, Kun; Liu, Zhiyuan; Zhang, Cunfang; Xin, Ying; Zhang, Zhiying

2015-06-01

In addition to the advantages of scalable, affordable, and easy to engineer, the clustered regularly interspaced short palindromic repeats (CRISPR)/CRISPR-associated protein (Cas) technology is superior for multiplex targeting, which is laborious and inconvenient when achieved by cloning multiple gRNA expressing cassettes. Here, we report a simple CRISPR array assembling method which will facilitate multiplex targeting usage. First, the Streptococcus thermophilus CRISPR3/Cas locus was cloned. Second, different CRISPR arrays were assembled with different crRNA spacers. Transformation assays using different Escherichia coli strains demonstrated efficient plasmid DNA targeting, and we achieved targeting efficiency up to 95% with an assembled CRISPR array with three crRNA spacers. Copyright © 2015 Elsevier Inc. All rights reserved.

Results of post-irradiation examination of WWER fuel assembly structural components made of E110 and E635 alloys

International Nuclear Information System (INIS)

Smirnov, A.; Markov, D.; Smirnov, V.; Polenok, V.; Ivashchenko, A.; Strozhuk, A.

2006-01-01

The paper presents the main examination results on the condition of fuel rods claddings, guide tubes and spacer grids of the WWER FA made of E110 and E635 alloys operated under standard operating conditions. The paper is based on the data obtained during the examination of 28 WWER-1000 FA and 12 WWER-400 FA. E110 alloy is shown to be suitable material for the WWER fuel rod claddings under the normal operating conditions. E635 alloy is attractive to manufacturing of the skeleton components. The currently used combination (E110 as a material of fuel rods claddings and E635 - as a material of the skeleton components) is the optimal solution for the WWER fuel assembly because the advantages of the both alloys are used. (authors)

Improvements in nuclear fuel assembly cages

Energy Technology Data Exchange (ETDEWEB)

Eaton, C.W.; Seeley, T.A.; Ince, G.; Speakman, W.T.

1986-03-12

The fuel pin/guide tube supporting grids of an assembly cage for a multi pin fuel element or a reflector element for a stringer are mounted in the moderator sleeve by way of mounting assemblies engaged in grooves machined into the interior surface of the sleeve, each mounting assembly including a split ring which is assembled into its groove by being radially contracted, pushed along the sleeve into registry with the groove and allowed to radially expand. The split ring may carry burnable neutron absorber. The region of the sleeve between two adjacent grids may be of smaller internal diameter than the remainder of the sleeve.

Hierarchical recruitment of ribosomal proteins and assembly factors remodels nucleolar pre-60S ribosomes.

Science.gov (United States)

Biedka, Stephanie; Micic, Jelena; Wilson, Daniel; Brown, Hailey; Diorio-Toth, Luke; Woolford, John L

2018-04-24

Ribosome biogenesis involves numerous preribosomal RNA (pre-rRNA) processing events to remove internal and external transcribed spacer sequences, ultimately yielding three mature rRNAs. Removal of the internal transcribed spacer 2 spacer RNA is the final step in large subunit pre-rRNA processing and begins with endonucleolytic cleavage at the C 2 site of 27SB pre-rRNA. C 2 cleavage requires the hierarchical recruitment of 11 ribosomal proteins and 14 ribosome assembly factors. However, the function of these proteins in C 2 cleavage remained unclear. In this study, we have performed a detailed analysis of the effects of depleting proteins required for C 2 cleavage and interpreted these results using cryo-electron microscopy structures of assembling 60S subunits. This work revealed that these proteins are required for remodeling of several neighborhoods, including two major functional centers of the 60S subunit, suggesting that these remodeling events form a checkpoint leading to C 2 cleavage. Interestingly, when C 2 cleavage is directly blocked by depleting or inactivating the C 2 endonuclease, assembly progresses through all other subsequent steps. © 2018 Biedka et al.

High corrosion-resistant fuel spacers

International Nuclear Information System (INIS)

Yoshida, Toshimi; Takase, Iwao; Ikeda, Shinzo; Masaoka, Isao; Nakajima, Junjiro.

1986-01-01

Purpose: To enable manufacturing BWR fuel spacers by prior-art production process, using a zirconium-base alloy having very excellent corrosion resistance. Method: A highly improved nodular-resistant, corrosion-resistant zirconium alloy is devised by adding a slight amount of niobium, titanium and vanadium to zircaloy, of which fuel spacers are produced. That is, there can be obtained an alloy having much more excellent nodular resistance than conventional zircaloy, and free from a large change in plasticity, workability, and weldability, by adding to zirconium about 1.5 % of tin, about 0.15 % of iron, about 0.05 % of chromium, about 0.05 % of nickel, and 0.05 to 0.5 % of at least one or two kinds of niobium, titanium and vanadium. Using this zirconium-base alloy can manufacture fuel spacers by the same manufacturing process, thus improving economy and reliability. (Kamimura, M.)

Performance of the Westinghouse WWER-1000 fuel design

International Nuclear Information System (INIS)

Höglund, J.; Jansson, A.; Latorre, R.; Davis, D.

2015-01-01

2012, several WFAs experienced mechanical damage of the spacer grids due to high lateral loads during core loading of both SU2 and SU3. The high loads were caused by mechanical interference due to a combination of distortion and stiffness of the mixed core. However, it shall be noted that all fuel rods remained hermetically sealed and non-leaking. Moreover incremental grid damage has not been observed on any WFA in any consecutive outage. To prevent damage of the WFA spacer grids during core loading and unloading, Westinghouse modified the WWER-1000 fuel design further to increase lateral grid strength and to minimize the risk for harmful mechanical interaction between assemblies. The design includes a thicker spacer grid outer strap with an enhanced profile, an all Alloy 718 grid structure for improved stability, and improvements to the top and bottom nozzles. (authors)

Preliminary thermal analysis of grids for twin source extraction system

International Nuclear Information System (INIS)

Pandey, Ravi; Bandyopadhyay, Mainak; Chakraborty, Arun K.

2017-01-01

The TWIN (Two driver based Indigenously built Negative ion source) source provides a bridge between the operational single driver based negative ion source test facility, ROBIN in IPR and an ITER-type multi driver based ion source. The source is designed to be operated in CW mode with 180kW, 1MHz, 5s ON/600s OFF duty cycle and also in 5Hz modulation mode with 3s ON/20s OFF duty cycle for 3 such cycle. TWIN source comprises of ion source sub-assembly (consist of driver and plasma box) and extraction system sub-assembly. Extraction system consists of Plasma grid (PG), extraction grid (EG) and Ground grid (GG) sub assembly. Negative ion beams produced at plasma grid seeing the plasma side of ion source will receive moderate heat flux whereas the extraction grid and ground grid would be receiving majority of heat flux from extracted negative ion and co-extracted electron beams. Entire Co-extracted electron beam would be dumped at extraction grid via electron deflection magnetic field making the requirement of thermal and hydraulic design for extraction grid to be critical. All the three grids are made of OFHC Copper and would be actively water cooled keeping the peak temperature rise of grid surface within allowable limit with optimum uniformity. All the grids are to be made by vacuum brazing process where joint strength becomes crucial at elevated temperature. Hydraulic design must maintain the peak temperature at the brazing joint within acceptable limit

Fuel sub-assembly

International Nuclear Information System (INIS)

Jolly, R.

1982-01-01

A fuel sub-assembly for a liquid metal cooled nuclear reactor is described in which the bundle of fuel pins are braced apart by a series of spaced grids. The grids at the lower end are capable of yielding, thus allowing pins swollen by irradiation to be withdrawn with a reduced risk of damage. (U.K.)

Nanoparticle-Based Brachytherapy Spacers for Delivery of Localized Combined Chemoradiation Therapy

Energy Technology Data Exchange (ETDEWEB)

Kumar, Rajiv, E-mail: r.kumar@neu.edu [Nanomedicine Science and Technology Center, Northeastern University, Boston, Massachusetts (United States); Department of Radiation Oncology, Brigham and Women' s Hospital, Dana-Farber Cancer Institute and Harvard Medical School, Boston, Massachusetts (United States); Belz, Jodi [Nanomedicine Science and Technology Center, Northeastern University, Boston, Massachusetts (United States); Markovic, Stacey [Department of Electrical and Computer Engineering, Northeastern University, Boston, Massachusetts (United States); Jadhav, Tej; Fowle, William [Nanomedicine Science and Technology Center, Northeastern University, Boston, Massachusetts (United States); Niedre, Mark [Department of Electrical and Computer Engineering, Northeastern University, Boston, Massachusetts (United States); Cormack, Robert; Makrigiorgos, Mike G. [Department of Radiation Oncology, Brigham and Women' s Hospital, Dana-Farber Cancer Institute and Harvard Medical School, Boston, Massachusetts (United States); Sridhar, Srinivas [Nanomedicine Science and Technology Center, Northeastern University, Boston, Massachusetts (United States); Department of Radiation Oncology, Brigham and Women' s Hospital, Dana-Farber Cancer Institute and Harvard Medical School, Boston, Massachusetts (United States)

2015-02-01

Purpose: In radiation therapy (RT), brachytherapy-inert source spacers are commonly used in clinical practice to achieve high spatial accuracy. These implanted devices are critical technical components of precise radiation delivery but provide no direct therapeutic benefits. Methods and Materials: Here we have fabricated implantable nanoplatforms or chemoradiation therapy (INCeRT) spacers loaded with silica nanoparticles (SNPs) conjugated containing a drug, to act as a slow-release drug depot for simultaneous localized chemoradiation therapy. The spacers are made of poly(lactic-co-glycolic) acid (PLGA) as matrix and are physically identical in size to the commercially available brachytherapy spacers (5 mm × 0.8 mm). The silica nanoparticles, 250 nm in diameter, were conjugated with near infrared fluorophore Cy7.5 as a model drug, and the INCeRT spacers were characterized in terms of size, morphology, and composition using different instrumentation techniques. The spacers were further doped with an anticancer drug, docetaxel. We evaluated the in vivo stability, biocompatibility, and biodegradation of these spacers in live mouse tissues. Results: The electron microscopy studies showed that nanoparticles were distributed throughout the spacers. These INCeRT spacers remained stable and can be tracked by the use of optical fluorescence. In vivo optical imaging studies showed a slow diffusion of nanoparticles from the spacer to the adjacent tissue in contrast to the control Cy7.5-PLGA spacer, which showed rapid disintegration in a few days with a burst release of Cy7.5. The docetaxel spacers showed suppression of tumor growth in contrast to control mice over 16 days. Conclusions: The imaging with the Cy7.5 spacer and therapeutic efficacy with docetaxel spacers supports the hypothesis that INCeRT spacers can be used for delivering the drugs in a slow, sustained manner in conjunction with brachytherapy, in contrast to the rapid clearance of the drugs when

Nanoparticle-Based Brachytherapy Spacers for Delivery of Localized Combined Chemoradiation Therapy

International Nuclear Information System (INIS)

Kumar, Rajiv; Belz, Jodi; Markovic, Stacey; Jadhav, Tej; Fowle, William; Niedre, Mark; Cormack, Robert; Makrigiorgos, Mike G.; Sridhar, Srinivas

2015-01-01

Purpose: In radiation therapy (RT), brachytherapy-inert source spacers are commonly used in clinical practice to achieve high spatial accuracy. These implanted devices are critical technical components of precise radiation delivery but provide no direct therapeutic benefits. Methods and Materials: Here we have fabricated implantable nanoplatforms or chemoradiation therapy (INCeRT) spacers loaded with silica nanoparticles (SNPs) conjugated containing a drug, to act as a slow-release drug depot for simultaneous localized chemoradiation therapy. The spacers are made of poly(lactic-co-glycolic) acid (PLGA) as matrix and are physically identical in size to the commercially available brachytherapy spacers (5 mm × 0.8 mm). The silica nanoparticles, 250 nm in diameter, were conjugated with near infrared fluorophore Cy7.5 as a model drug, and the INCeRT spacers were characterized in terms of size, morphology, and composition using different instrumentation techniques. The spacers were further doped with an anticancer drug, docetaxel. We evaluated the in vivo stability, biocompatibility, and biodegradation of these spacers in live mouse tissues. Results: The electron microscopy studies showed that nanoparticles were distributed throughout the spacers. These INCeRT spacers remained stable and can be tracked by the use of optical fluorescence. In vivo optical imaging studies showed a slow diffusion of nanoparticles from the spacer to the adjacent tissue in contrast to the control Cy7.5-PLGA spacer, which showed rapid disintegration in a few days with a burst release of Cy7.5. The docetaxel spacers showed suppression of tumor growth in contrast to control mice over 16 days. Conclusions: The imaging with the Cy7.5 spacer and therapeutic efficacy with docetaxel spacers supports the hypothesis that INCeRT spacers can be used for delivering the drugs in a slow, sustained manner in conjunction with brachytherapy, in contrast to the rapid clearance of the drugs when

Cold-neutron tomography of annular flow and functional spacer performance in a model of a boiling water reactor fuel rod bundle

International Nuclear Information System (INIS)

Zboray, Robert; Kickhofel, John; Damsohn, Manuel; Prasser, Horst-Michael

2011-01-01

Highlights: → Annular flows w/wo functional spacers are investigated by cold neutron imaging. → Liquid film thickness distribution on fuel pins and on spacer vanes is measured. → The influence of the spacers on the liquid film distributions has been quantified. → The cross-sectional averaged liquid hold-up significantly affected by the spacers. → The sapers affect the fraction of the entrained liquid hold up in the gas core. - Abstract: Dryout of the coolant liquid film at the upper part of the fuel pins of a boiling water reactor (BWR) core constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is both a safety concern and a limiting factor in the thermal power and thus for the economy of BWRs. We have investigated adiabatic, air-water annular flows in a scaled-up model of two neighboring subchannels as found in BWR fuel assemblies using cold-neutron tomography. The imaging of the double suchannel has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institute, Switzerland. Cold-neutron tomography is shown here to be an excellent tool for investigating air-water annular flows and the influence of functional spacers of different geometries on such flows. The high-resolution, high-contrast measurements provide the spatial distributions of the coolant liquid film thickness on the fuel pin surfaces as well as on the surfaces of the spacer vanes. The axial variations of the cross-section averaged liquid hold-up and its fraction in the gas core shows the effect of the spacers on the redistribution of the two phases.

Preliminary performance analysis of exponential experimental system for the determination of neutron effective multiplication factor of PWR spent fuel

International Nuclear Information System (INIS)

Shin, Heesung; Lee, Sang-Yun; Ro, Seung-Gy; Seo, Gi-Seok; Kim, Ho-Dong

2002-01-01

An exponential experiment system which is composed of neutron detector, signal analysis system and neutron source, 10 mCi Cf-252 has been installed in the storage pool of PIEF at KAERI in order to experimentally determining neutron effective multiplication factors of PWR spent fuel assemblies. Preliminary functional characteristic tests of the experimental system are performed for C15, J14 and J44 assemblies loaded in the pool. As a result of preliminary tests, the average neutron counts obtained for 3 minutes in the plateau of the C15, J14 and J44 assemblies are about 1900, 3800 and 3200, respectively. A dip of the neutron flux density distribution is noticed in the spacer grid position. Neutron counts at those positions appear to be reduced to about 70 % in comparison to the fuel position. The measured axial neutron distribution shapes are compared with the result for the P14 assembly and Cs-137 gamma scanning data performed in KAERI. It is revealed that the spacer grid position measured is consistent with the design specifications within a 2.3 % error. The exponential decay constants for the C15 assembly were determined to be 0.152 and 0.165 for detector and source scanning, respectively. (author)

A Study on the Fuel Assembly Seismic Analysis without Holddown Springs

International Nuclear Information System (INIS)

Kwon, O Cheol; Ha, Dong Geun; Lee, Kyou Seok; Jeon, Sang Yoon; Suh, Jung Min

2013-01-01

In this study, the effect for the fuel assembly removed holddown spring under seismic event has been evaluated through the comparison with the seismic analysis result of fuel assembly with holddown spring. In order to compare each design, the simplified fuel assembly seismic analysis models have been established according to reference. The mid grid impact force, natural frequency, and top nozzle displacement for each fuel assembly model has been analyzed using ANSYS. The fuel assembly seismic analyses without holddown springs are performed and compared to the model with holddown springs. The grid impact forces of CPM 1 and CPM 2 are almost doubled in comparison with CPM 3 and almost tripled in comparison with CPM 4 so the grid impact forces depend on CPM types. The grid impact forces of the fuel assembly model without holddown springs have similar tendencies in comparison with fuel assembly with holddown springs. Moreover, the model without holddown springs analysis time is much longer than the model with holddown springs. Consequently, it is moderate that the fuel assembly analysis model with holddown springs would be used for effective analysis even though the actual model has no holddown springs

A Study on the Fuel Assembly Seismic Analysis without Holddown Springs

Energy Technology Data Exchange (ETDEWEB)

Kwon, O Cheol; Ha, Dong Geun; Lee, Kyou Seok; Jeon, Sang Yoon; Suh, Jung Min [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of)

2013-10-15

In this study, the effect for the fuel assembly removed holddown spring under seismic event has been evaluated through the comparison with the seismic analysis result of fuel assembly with holddown spring. In order to compare each design, the simplified fuel assembly seismic analysis models have been established according to reference. The mid grid impact force, natural frequency, and top nozzle displacement for each fuel assembly model has been analyzed using ANSYS. The fuel assembly seismic analyses without holddown springs are performed and compared to the model with holddown springs. The grid impact forces of CPM{sub 1} and CPM{sub 2} are almost doubled in comparison with CPM{sub 3} and almost tripled in comparison with CPM{sub 4} so the grid impact forces depend on CPM types. The grid impact forces of the fuel assembly model without holddown springs have similar tendencies in comparison with fuel assembly with holddown springs. Moreover, the model without holddown springs analysis time is much longer than the model with holddown springs. Consequently, it is moderate that the fuel assembly analysis model with holddown springs would be used for effective analysis even though the actual model has no holddown springs.

Sub-channel analysis of a HPLWR fuel assembly with STAR-CD

International Nuclear Information System (INIS)

Himmel, Steffen R.; Class, Andreas G.; Schulenberg, Thomas; Laurien, Eckart

2008-01-01

Hofmeister et. al. developed a first design proposal for a HPLWR fuel assembly, consisting of a square 7 by 7 fuel pin arrangement within an assembly box and a water box in the centre, replacing 9 fuel rods. Instead of conventional grid spacers, wire wraps are considered due to good coolant mixing and low pressure drop in either flow direction. Within the present work, a novel approach describing the coolant heat up in the sub-channels of such an assembly has been investigated: the commercial software package STAR-CD has been used as a sub-channel code to investigate the thermal-hydraulic performance of such an HPLWR fuel assembly. The aim of the work is to demonstrate that a widely accepted commercial Computational Fluid Dynamics (CFD) code can be used for full rod bundle analysis by applying minor modifications to it. In steady of writing a dedicated code system with numerical solver routines and post-processing tools for sub-channel analyses, the user benefits from the optimized Graphical User Interface (GUI) already provided in STAR-CD. Moreover, a smooth transition to full three-dimensional modeling of the fluid flow inside rod bundles will be possible with the same code system, if considered to be necessary, just by refining the spatial discretization. Steady-state and transient flow regimes can be studied for design as well as reactor safety analysis. As the STAR-CD code uses the Finite Volume Method (FVM) for spatial discretization, the conservation equations for mass, momentum and energy were modified via user-subroutines to obtain the equations known from the usual sub-channel approach. The method will be explained in detail and results will be discussed. (author)

Nuclear fuel assembly

International Nuclear Information System (INIS)

Delafosse, Jacques.

1977-01-01

This invention relates to a nuclear fuel assembly for a light or heavy water reactor, or for a fast reactor of the kind with a bundle of cladded pins, maintained parallel to each other in a regular network by an assembly of separate supporting grids, fitted with elastic bearing surfaces on these pins [fr

Pattern fuel assembly loading system

International Nuclear Information System (INIS)

Ahmed, H.J.; Gerkey, K.S.; Miller, T.W.; Wylie, M.E.

1986-01-01

This patent describes an interactive system for facilitating preloading of fuel rods into magazines, which comprises: an operator work station adapted for positioning between a supply of fuel rods of predetermined types, and the magazine defining grid locations for a predetermined fuel assembly; display means associated with the work station; scanner means associated with the work station and adapted for reading predetermined information accompanying the fuel rods; a rectangular frame adapted for attachment to one end of the fuel assembly loading magazine; prompter/detector means associated with the frame for detecting insertion of a fuel rod into the magazine; and processing means responsive to the scanner means and the sensing means for prompting the operator via the display means to pre-load the fuel rods into desired grid locations in the magazine. An apparatus is described for facilitating pre-loading of fuel rods in predetermined grid locations of a fuel assembly loading magazine, comprising: a rectangular frame adapted for attachment to one end of the fuel assembly loading magazine; and means associated with the frame for detecting insertion of fuel rods into the magazine

Process and installation for welding nuclear fuel assembly grids

International Nuclear Information System (INIS)

Vere, B.; Mathevon, P.

1985-01-01

The invention proposes a process to weld two sets of perpendicular plates of which the end parts are made integral with a belt piece; the grid is held in a support frame with access openings to the points to be welded on the two faces and on the grid sides; the frame is moved on a mobile table by means of an orientation system along the perpendicular direction of an electron beam welding equipment; each joint to be welded is presented, rotating the frame through 90 deg about an axis and repeating the operation, and rotating the frame about a perpendicular axis and repeating the operation until all the joints on each side of the grid have been welded [fr

Preclinical Evaluation of Bioabsorbable Polyglycolic Acid Spacer for Particle Therapy

International Nuclear Information System (INIS)

Akasaka, Hiroaki; Sasaki, Ryohei; Miyawaki, Daisuke; Mukumoto, Naritoshi; Sulaiman, Nor Shazrina Binti; Nagata, Masaaki; Yamada, Shigeru; Murakami, Masao; Demizu, Yusuke; Fukumoto, Takumi

2014-01-01

Purpose: To evaluate the efficacy and safety of a polyglycolic acid (PGA) spacer through physical and animal experiments. Methods and Materials: The spacer was produced with surgical suture material made of PGA, forming a 3-dimensional nonwoven fabric. For evaluation or physical experiments, 150-MeV proton or 320-MeV carbon-ion beams were used to generate 60-mm width of spread-out Bragg peak. For animal experiments, the abdomens of C57BL/6 mice, with or without the inserted PGA spacers, were irradiated with 20 Gy of carbon-ion beam (290 MeV) using the spread-out Bragg peak. Body weight changes over time were scored, and radiation damage to the intestine was investigated using hematoxylin and eosin stain. Blood samples were also evaluated 24 days after the irradiation. Long-term thickness retention and safety were evaluated using crab-eating macaques. Results: No chemical or structural changes after 100 Gy of proton or carbon-ion irradiation were observed in the PGA spacer. Water equivalency of the PGA spacer was equal to the water thickness under wet condition. During 24 days' observation after 20 Gy of carbon-ion irradiation, the body weights of mice with the PGA spacer were relatively unchanged, whereas significant weight loss was observed in those mice without the PGA spacer (P<.05). In mice with the PGA spacer, villus and crypt structure were preserved after irradiation. No inflammatory reactions or liver or renal dysfunctions due to placement of the PGA spacer were observed. In the abdomen of crab-eating macaques, thickness of the PGA spacer was maintained 8 weeks after placement. Conclusions: The absorbable PGA spacer had water-equivalent, bio-compatible, and thickness-retaining properties. Although further evaluation is warranted in a clinical setting, the PGA spacer may be effective to stop proton or carbon-ion beams and to separate normal tissues from the radiation field

Oriented crystalline monolayers and bilayers of 2 x 2 silver(1) grid architectures at the air-solution interface: Their assembly and crystal structure elucidation

DEFF Research Database (Denmark)

Weissbuch, J.; Baxter, P.N.W.; Kuzmenko, I.

2000-01-01

Oriented crystalline monolayers, similar to 14 Angstrom thick, of a 2 x 2 Ag+ grid complex, self-assembled at the air-solution interface starting from an water-insoluble ligand 3,6-bis[2-(6-phenylpyridine)]pyridazine spread on silver-ion-containing solutions,were examined by grazing-incidence X...

Bioinformatics analyses of Shigella CRISPR structure and spacer classification.

Science.gov (United States)

Wang, Pengfei; Zhang, Bing; Duan, Guangcai; Wang, Yingfang; Hong, Lijuan; Wang, Linlin; Guo, Xiangjiao; Xi, Yuanlin; Yang, Haiyan

2016-03-01

Clustered regularly interspaced short palindromic repeats (CRISPR) are inheritable genetic elements of a variety of archaea and bacteria and indicative of the bacterial ecological adaptation, conferring acquired immunity against invading foreign nucleic acids. Shigella is an important pathogen for anthroponosis. This study aimed to analyze the features of Shigella CRISPR structure and classify the spacers through bioinformatics approach. Among 107 Shigella, 434 CRISPR structure loci were identified with two to seven loci in different strains. CRISPR-Q1, CRISPR-Q4 and CRISPR-Q5 were widely distributed in Shigella strains. Comparison of the first and last repeats of CRISPR1, CRISPR2 and CRISPR3 revealed several base variants and different stem-loop structures. A total of 259 cas genes were found among these 107 Shigella strains. The cas gene deletions were discovered in 88 strains. However, there is one strain that does not contain cas gene. Intact clusters of cas genes were found in 19 strains. From comprehensive analysis of sequence signature and BLAST and CRISPRTarget score, the 708 spacers were classified into three subtypes: Type I, Type II and Type III. Of them, Type I spacer referred to those linked with one gene segment, Type II spacer linked with two or more different gene segments, and Type III spacer undefined. This study examined the diversity of CRISPR/cas system in Shigella strains, demonstrated the main features of CRISPR structure and spacer classification, which provided critical information for elucidation of the mechanisms of spacer formation and exploration of the role the spacers play in the function of the CRISPR/cas system.

Biomechanical Analysis of an Expandable Lumbar Interbody Spacer.

Science.gov (United States)

Soriano-Baron, Hector; Newcomb, Anna G U S; Malhotra, Devika; Palma, Atilio E; Martinez-Del-Campo, Eduardo; Crawford, Neil R; Theodore, Nicholas; Kelly, Brian P; Kaibara, Taro

2018-06-01

Recently developed expandable interbody spacers are widely accepted in spinal surgery; however, the resulting biomechanical effects of their use have not yet been fully studied. We analyzed the biomechanical effects of an expandable polyetheretherketone interbody spacer inserted through a bilateral posterior approach with and without different modalities of posterior augmentation. Biomechanical nondestructive flexibility testing was performed in 7 human cadaveric lumbar (L2-L5) specimens followed by axial compressive loading. Each specimen was tested under 6 conditions: 1) intact, 2) bilateral L3-L4 cortical screw/rod (CSR) alone, 3) WaveD alone, 4) WaveD + CSR, 5) WaveD + bilateral L3-L4 pedicle screw/rod (PSR), and 6) WaveD + CSR/PSR, where CSR/PSR was a hybrid construct comprising bilateral cortical-level L3 and pedicle-level L4 screws interconnected by rods. The range of motion (ROM) with the interbody spacer alone decreased significantly compared with the intact condition during flexion-extension (P = 0.02) but not during lateral bending or axial rotation (P ≥ 0.19). The addition of CSR or PSR to the interbody spacer alone condition significantly decreased the ROM compared with the interbody spacer alone (P ≤ 0.002); and WaveD + CSR, WaveD + PSR, and WaveD + CSR/PSR (hybrid) (P ≥ 0.29) did not differ. The axial compressive stiffness (resistance to change in foraminal height during compressive loading) with the interbody spacer alone did not differ from the intact condition (P = 0.96), whereas WaveD + posterior instrumentation significantly increased compressive stiffness compared with the intact condition and the interbody spacer alone (P ≤ 0.001). The WaveD alone significantly reduced ROM during flexion-extension while maintaining the axial compressive stiffness. CSR, PSR, and CSR/PSR hybrid constructs were all effective in augmenting the expandable interbody spacer system and improving its stability. Copyright © 2018 Elsevier Inc. All

Design requirement on HYPER blanket fuel assembly

International Nuclear Information System (INIS)

Hwang, Woan; Lee, B. O.; Nam, C.; Ryu, W. S.; Lee, B. S.; Park, W. S.

2000-07-01

This document describes design requirements which are needed for designing the blanket assembly of the HYPER as design guidance. The blanket assembly of the HYPER consists of blanket fuel rods, mounting rail, spacer, upper nozzle with handling socket, bottom nozzle with mounting rail and skeleton structure. The blanket fuel rod consists of top end plug, bottom end plug with key way, blanket fuel slug, and cladding. In the assembly, the rods are in a triangular pitch array. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements for the blanket fuel assembly of the HYPER

Grid sleeve bulge tool

International Nuclear Information System (INIS)

Phillips, W.D.; Vaill, R.E.

1980-01-01

An improved grid sleeve bulge tool is designed for securing control rod guide tubes to sleeves brazed in a fuel assembly grid. The tool includes a cylinder having an outer diameter less than the internal diameter of the control rod guide tubes. The walls of the cylinder are cut in an axial direction along its length to provide several flexible tines or ligaments. These tines are similar to a fork except they are spaced in a circumferential direction. The end of each alternate tine is equipped with a semispherical projection which extends radially outwardly from the tine surface. A ram or plunger of generally cylindrical configuration and about the same length as the cylinder is designed to fit in and move axially of the cylinder and thereby force the tined projections outwardly when the ram is pulled into the cylinder. The ram surface includes axially extending grooves and plane surfaces which are complimentary to the inner surfaces formed on the tines on the cylinder. As the cylinder is inserted into a control rod guide tube, and the projections on the cylinder placed in a position just below or above a grid strap, the ram is pulled into the cylinder, thus moving the tines and the projections thereon outwardly into contact with the sleeve, to plastically deform both the sleeve and the control rod guide tube, and thereby form four bulges which extend outwardly from the sleeve surface and beyond the outer periphery of the grid peripheral strap. This process is then repeated at the points above the grid to also provide for outwardly projecting surfaces, the result being that the grid is accurately positioned on and mechanically secured to the control rod guide tubes which extend the length of a fuel assembly

Grid for nuclear fuel rod assembly

International Nuclear Information System (INIS)

Brayman, K.W.; George, D.K.; Rawlings, J.C.; Dix, G.E.

1976-01-01

A grid is described for placing a least four corner fuel rods in a tubular flow channel of a nuclear reactor. It includes a bearer component composed of four side strips joined by four corner strips so as to form a rigid unit structure, each side strip having an L-shaped piece adjacent at each of its ends to a lug of each L-shaped piece extending to the adjacent end of its associated side strip [fr

Development of guide thimble stress peaking factor calculation methodology

International Nuclear Information System (INIS)

Lee, Seong Ki; Jeon, Sang Youn; Kim, Jae Ik; Jeon, Kyeong Lak; Kim, Kyu Tae

2004-01-01

The Nuclear Fuel Assembly for light water reactor which provides for 236 fuel rods consists of guide tubes, spacer grids, top/bottom nozzles. The guide tubes form the main structural components in conjunction with the grids, act as the main load carrying members of fuel assembly and serve as a support structure and a guide path for the control element, neutron sources and incore instruments after they are secured to upper and lower end areas. Top/bottom nozzles make the end parts of fuel assembly. And the spacer girds maintain the fuel rod array by providing positive lateral restraint to the fuel rod to the fuel rod but only frictional restraint to axial fuel rod motion. When the fuel assembly is in reactor, the tensional and compressional forces are applied to guide thimble through the top nozzle. The stresses vary with the location of guide thimble on the top nozzle plate since the different flow plate thickness between center and outer areas causes a different flexibility. The relative stress shall be considered during designing this kind of structure. And it is useful to know a coefficient to represent this relative stress difference and this value is called stress peaking factor

PCR-Internal Transcribed Spacer (ITS) genes sequencing and ...

African Journals Online (AJOL)

Methods: DNA extraction, purification, amplification and sequencing of Internal Transcribed Spacer (ITS) genes were per- formed using ... Keywords: Internal transcribed spacer genes, phylogenetic, genetic relationship, clinical and environmental fungi, HIV-TB. ... Nigeria. An Ethical clearance was obtained from the Eth-.

Irradiation of Parts of the X-Gen Nuclear Fuel Assembly made by KNF in HANARO

International Nuclear Information System (INIS)

Choo, K. N.; Cho, M. S.; Shin, Y. T.; Kim, B. G.; Lee, S. H.; Eom, K. B.

2008-01-01

An instrumented capsule has been developed at HANARO (High flux Advanced Neutron Application ReactOr) for the neutron irradiation tests of materials. The capsule system has been actively utilized for the various material irradiation tests requested by users from research institutes, universities, and the industries. As a preliminary test, some specimens made of the parts of a nuclear fuel assembly were inserted in the 05M-07U instrumented capsule and successfully irradiated at HANARO. Based on the results and experience, a new irradiation capsule of 07M-13N was designed, fabricated, and irradiated at HANARO for the evaluation of the neutron irradiation properties of the parts of the X-Gen nuclear fuel assembly made by KNF (Korea Nuclear Fuel). Specimens such as bucking and spring test specimens of spacer grid, microstructure and tensile test specimens of welded parts, tensile, irradiation growth and spring test specimens made of HANA tube, Zirlo, Zircaloy-4 and Inconel-718 were placed in the capsule. The capsule was loaded into the CT test hole of HANARO of a 30MW thermal output and the specimens were irradiated at 295 - 460 .deg. C up to a fast neutron fluence of 1.2x10 21 (n/cm 2 ) (E>1.0MeV)

Cell for receipting and dismantling nuclear fuel assembly

International Nuclear Information System (INIS)

Beneck, J.A.; Quayre, C.

1989-01-01

The cell has a vertical structure with a right section corresponding at that of the assembly to receive, a mechanism for keeping fuel pins at their nominal separation in the form of at least two combs and mechanisms of holding grids and bottom nozzle. The comb arrangements are moved into position by hydraulic actuators so that they cross each other to form a lattice round the fuel pins. The mechanism for holding grid assemblies consist of joints that articulate from a free position to a position where the joints press of the grid on all sides [fr

Shock absorbing structure for nuclear fuel assemblies

International Nuclear Information System (INIS)

Jabsen, F.S.

1981-01-01

A hydraulic apparatus is described that absorbs shocks that may be applied to fuel assemblies. Spring pads mounted on the upper end fittings of the fuel assemblies have plungers that move within hollow guide posts attached to the upper grids of the fuel assemblies. (L.L.)

Membrane with integrated spacer

NARCIS (Netherlands)

Balster, J.H.; Stamatialis, Dimitrios; Wessling, Matthias

2010-01-01

Many membrane processes are severely influenced by concentration polarisation. Turbulence promoting spacers placed in between the membranes can reduce the diffusional resistance of concentration polarisation by inducing additional mixing. Electrodialysis (ED) used for desalination suffers from

A CAREM fuel assembly prototype construction in order to verify its mechanical design using hydrodynamic testing

International Nuclear Information System (INIS)

Aparicio, Gaspar; Di Marco, Agustin; Falcone, Jose M.; Giorgis, Miguel A.; Mathot, Sergio R.; Migliori, Julio; Orlando, Oscar S.; Restelli, Miguel A.; Ruggirello, Gabriel; Sapia, Gustavo C.; Zinzallari, Fausto; Bianchi, Daniel R.; Volpi, Ricardo M.

2000-01-01

The scope of this paper is to describe the activities of several Groups from three Atomic Centers (C. A. Bariloche, C. A. Ezeiza and C. A. Constituyentes), involved in the manufacturing of a CAREM fuel assembly prototype. The Design Group (UAIN-CAB) carried out the fuel assembly engineering. Cladding components were constructed by the Special Alloys Pilot Factory (UAMCN-CAE). Engineering Group (UACN-CAC) manufactured the parts to be processed, resorting to qualified suppliers. Elastic spacers were completely designed and constructed by this Group, and fuel rods, control rods, guide tubes and spacers were also welded here. Research Reactors Fuels Group (UACN-CAC) carried out the dimensional control of the elaborated parts, while Postirradiation Testing Group (UACN-CAC) performed the assembling of the fuel element. This paper also refers to the design and development of special equipment and devices, all of them required for the prototype construction. (author)

Nuclear fuel assembly grid sleeve/guide thimble bulge orientation gage and inspection method

International Nuclear Information System (INIS)

Widener, W.H.

1988-01-01

This patent describes a method of inspecting a fuel assembly to determine the orientation of externally-projecting mated bulges connecting a grid sleeve to a guide thimble of the assembly, the method comprising the steps of: (a) inserting a radially-expandable tubular member within the guide thimble, the tubular member having externally-projecting embossments thereon spaced circumferentially from one another about the tubular member, the embossments being the same in number as the bulges of the guide thimble and configured to fit therewithin; (b) axially moving an elongated expansion member, which extends through and rotatably mounts the tubular member, relative to the tubular member from a first position in which the expansion member permits inward contraction of the tubular member and displacement of embossments thereon away from the interior of the guide thimble bulges for removing the embossments from registry therewith and a second position in which the expansion member produces radial expansion of the tubular member and displacement of the embossments thereon toward the interior of the guide thimble bulges for placing the embossments in registry therewith; (c) rotating the tubular member relative to the expansion member so as to bring the embossments on the tubular member into alignment with the guide thimble bulges as the embossments on the tubular member are being displaced toward and into registry with the interior of the bulges; and (d) responsive to rotation of the tubular member away from a reference position, providing an indication of the orientation of the guide thimble bulges relative to a reference point upon displacement of the embossments into registry therewith

Shape modification for decreasing the spring stiffness of double-plate nozzle type spacer grid spring

International Nuclear Information System (INIS)

Lee, K. H.; Kang, H. S.; Song, K. N.; Yun, K. H.; Kim, H. K.

2001-01-01

Nozzle of the double-plated grid plays the role of the spirng to support a fuel rod as well as the coolant path in grid. The nozzle was known to be necessary to reduce the spring stiffness for supporting performance. In this study, the contact analysis between the fuel rod and the newly designed nozzle was performed by ABAQUS computer code to propose the preferable shape in term of spring performance. Two small cut at the upper and lower part of the nozzle appeared to have a minor effect in decreasing the nozzle stiffness. A long slot at the center of the nozzle was turned out not only to decrease the spring constant as desired but also to increase the elastic displacement

In vitro microbiologic evaluation of PTFE and cotton as spacer materials.

Science.gov (United States)

Paranjpe, Avina; Jain, Sumita; Alibhai, Karim J; Wadhwani, Chandur P; Darveau, Richard P; Johnson, James D

2012-09-01

To microbiologically evaluate the efficacy of cotton and polytetrafluoroethylene (PTFE) tape used as spacer materials. Twenty-six extracted human molars were restored using either cotton or PTFE tape as spacers under a standardized provisional restorative material (Cavit). The teeth were incubated for 7 days in a culture of Streptococcus gordonii or in liquid media alone. The spacers were removed and tested for bacterial contamination. The access cavities were also evaluated for bacterial contamination. Nine of 10 teeth with cotton spacers and one of 10 teeth with PTFE spacers were positive for S gordonii growth. The nine teeth in the cotton group also showed contamination of the access cavities. Even under optimal conditions, cotton spacers may cause leakage into the access cavities. Cotton fibers may serve as a route for bacterial contamination of the access cavities and root canal space. In contrast, PTFE tape did not provide an avenue for bacterial contamination.

The CRISPR Spacer Space Is Dominated by Sequences from Species-Specific Mobilomes.

Science.gov (United States)

Shmakov, Sergey A; Sitnik, Vassilii; Makarova, Kira S; Wolf, Yuri I; Severinov, Konstantin V; Koonin, Eugene V

2017-09-19

Clustered regularly interspaced short palindromic repeats and CRISPR-associated protein (CRISPR-Cas) systems store the memory of past encounters with foreign DNA in unique spacers that are inserted between direct repeats in CRISPR arrays. For only a small fraction of the spacers, homologous sequences, called protospacers, are detectable in viral, plasmid, and microbial genomes. The rest of the spacers remain the CRISPR "dark matter." We performed a comprehensive analysis of the spacers from all CRISPR- cas loci identified in bacterial and archaeal genomes, and we found that, depending on the CRISPR-Cas subtype and the prokaryotic phylum, protospacers were detectable for 1% to about 19% of the spacers (~7% global average). Among the detected protospacers, the majority, typically 80 to 90%, originated from viral genomes, including proviruses, and among the rest, the most common source was genes that are integrated into microbial chromosomes but are involved in plasmid conjugation or replication. Thus, almost all spacers with identifiable protospacers target mobile genetic elements (MGE). The GC content, as well as dinucleotide and tetranucleotide compositions, of microbial genomes, their spacer complements, and the cognate viral genomes showed a nearly perfect correlation and were almost identical. Given the near absence of self-targeting spacers, these findings are most compatible with the possibility that the spacers, including the dark matter, are derived almost completely from the species-specific microbial mobilomes. IMPORTANCE The principal function of CRISPR-Cas systems is thought to be protection of bacteria and archaea against viruses and other parasitic genetic elements. The CRISPR defense function is mediated by sequences from parasitic elements, known as spacers, that are inserted into CRISPR arrays and then transcribed and employed as guides to identify and inactivate the cognate parasitic genomes. However, only a small fraction of the CRISPR spacers

Pre-test nondestructive examination data summary report on Turkey Point spent fuel assemblies D01, D04 and D06 for the climax-spent fuel test

International Nuclear Information System (INIS)

Davis, R.B.

1981-01-01

Fuel assembly sip testing conducted at Turkey Point and Battelle Columbus Laboratories (BCL) confirmed no leaking rods were among the thirteen fuel assemblies included in the Climax-Spent Fuel Test. A detailed nondestructive examination was conducted on three of the thirteen assemblies. Fuel assembly lengths and widths averaged 153.6 inches and 8.3 inches, respectively. The assemblies weighed 1459 +- 3 lbs. Total neutron flux measured at the fuel column midplane was 1.06 x 10 4 N/cm 2 /s with an average neutron energy of 1.4 MeV. Gamma dose rates were measured axially and vertically to the fuel column with maximum contact dose rate of 9.52 x 10 4 R/h. Twenty rods underwent detailed rod nondestructive examination. Rod lengths and weights averaged 152.5 inches and 6.82 lb, respectively. Spiral profilometry scans showed the maximum ovality for the twenty rods was 0.0105 inch with average rod diameters ranging from 0.4201 inch to 0.4211 inch. Extensive ridging from pellet cladding interaction was evident over most of the length on all rods. Gamma scan results showed no cesium peaking and no unusually large pellet to pellet gaps. Approximate 10% gamma activity depressions were found at the grid spacer locations. Several areas were identified as locations with an internal anomaly using eddy current results. Fifteen rods were reinserted into the three fuel assemblies at the completion of the nondestructive examinations. Five rods remained at BCL for destructive characterization

Ion mobility spectrometer with virtual aperture grid

Science.gov (United States)

Pfeifer, Kent B.; Rumpf, Arthur N.

2010-11-23

An ion mobility spectrometer does not require a physical aperture grid to prevent premature ion detector response. The last electrodes adjacent to the ion collector (typically the last four or five) have an electrode pitch that is less than the width of the ion swarm and each of the adjacent electrodes is connected to a source of free charge, thereby providing a virtual aperture grid at the end of the drift region that shields the ion collector from the mirror current of the approaching ion swarm. The virtual aperture grid is less complex in assembly and function and is less sensitive to vibrations than the physical aperture grid.

Antibiotic-impregnated articulating cement spacer for infected total knee arthroplasty

Directory of Open Access Journals (Sweden)

Parag Garg

2011-01-01

Full Text Available Background : Standard treatment of chronic infected total knee arthroplasty (TKA is a two-stage revision, the first step being placement of an antibiotic-impregnated cement spacer. Here we describe the results of a new technique (modification of the Goldstien′s technique for intraoperative manufacture of a customized articulating spacer at minimal cost and with relatively good conformity and longevity. Materials and Methods : Thirty-six infected knees underwent this procedure from June 2002 to May 2007. The technique consists of using the freshened femur and tibia interface as molds wrapped in a tin foil for manufacturing the two components of the spacer with antibiotic-impregnated methyl methycrylate cement. We used the spacer and the femoral component of the trial set of a TKA system to mold them to perfect articulation. We also reinforced the spacer with a K-wire scaffold to prevent fracture of the cement mantle in the last 21 cases. Results : All 36 knees showed excellent results in terms of infection control, mobility, and stability. There was significant improvement in the WOMAC and Knee Society Scores (20 and 39 points respectively. There were two fractures of the spacers in the initial 15 cases that did not have K-wire scaffolding but none in the last 21 that had reinforcement. Conclusion : This technique provides a more conforming spacer, with good range of motion and stability. The reinforcement helps in preventing the fracture of the cement mantle and is cost effective.

Development and characterization of 3D-printed feed spacers for spiral wound membrane systems

KAUST Repository

Siddiqui, Amber

2016-01-02

Feed spacers are important for the impact of biofouling on the performance of spiral-wound reverse osmosis (RO) and nanofiltration (NF) membrane systems. The objective of this study was to propose a strategy for developing, characterizing, and testing of feed spacers by numerical modeling, three-dimensional (3D) printing of feed spacers and experimental membrane fouling simulator (MFS) studies. The results of numerical modeling on the hydraulic behavior of various feed spacer geometries suggested that the impact of spacers on hydraulics and biofouling can be improved. A good agreement was found for the modeled and measured relationship between linear flow velocity and pressure drop for feed spacers with the same geometry, indicating that modeling can serve as first step in spacer characterization. An experimental comparison study of a feed spacer currently applied in practice and a 3D printed feed spacer with the same geometry showed (i) similar hydraulic behavior, (ii) similar pressure drop development with time and (iii) similar biomass accumulation during MFS biofouling studies, indicating that 3D printing technology is an alternative strategy for development of thin feed spacers with a complex geometry. Based on the numerical modeling results, a modified feed spacer with low pressure drop was selected for 3D printing. The comparison study of the feed spacer from practice and the modified geometry 3D printed feed spacer established that the 3D printed spacer had (i) a lower pressure drop during hydraulic testing, (ii) a lower pressure drop increase in time with the same accumulated biomass amount, indicating that modifying feed spacer geometries can reduce the impact of accumulated biomass on membrane performance. The combination of numerical modeling of feed spacers and experimental testing of 3D printed feed spacers is a promising strategy (rapid, low cost and representative) to develop advanced feed spacers aiming to reduce the impact of biofilm formation on

(ISSR) and internal transcribed spacer

African Journals Online (AJOL)

Cladistic relationships within the genus Cinnamomum (Lauraceae) in Taiwan based on analysis of leaf morphology and inter-simple sequence repeat (ISSR) and internal transcribed spacer (ITS) molecular markers.

Reconstitutable fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Ferlan, S.J.; Kmonk, S.; Schallenberger, J.M.

1982-01-01

A reconstitutable fuel assembly for a nuclear reactor which includes a mechanical, rather than metallurgical, arrangement for connecting control rod guide thimbles to the top and bottom nozzles of a fuel assembly. Multiple sleeves enclosing control rod guide thimbles interconnect the top nozzle to the fuel assembly upper grid. Each sleeve is secured to the top nozzle by retaining rings disposed on opposite sides of the nozzle. Similar sleeves enclose the lower end of control rod guide thimbles and interconnect the bottom nozzle with the lowermost grid on the assembly. An end plug fitted in the bottom end of each sleeve extends through the bottom nozzle and is secured thereto by a retaining ring. Should it be necessary to remove a fuel rod from the assembly, the retaining rings in either the top or bottom nozzles may be removed to release the nozzle from the control rod guide thimbles and thus expose either the top or bottom ends of the fuel rods to fuel rod removing mechanisms

Orthognathic model surgery with LEGO key-spacer.

Science.gov (United States)

Tsang, Alfred Chee-Ching; Lee, Alfred Siu Hong; Li, Wai Keung

2013-12-01

A new technique of model surgery using LEGO plates as key-spacers is described. This technique requires less time to set up compared with the conventional plaster model method. It also retains the preoperative setup with the same set of models. Movement of the segments can be measured and examined in detail with LEGO key-spacers. Copyright © 2013 American Association of Oral and Maxillofacial Surgeons. Published by Elsevier Inc. All rights reserved.

Synthesis of the turbulent mixing in a rod bundle with vaned spacer grids based on the OECD-KAERI CFD benchmark exercise

International Nuclear Information System (INIS)

Lee, Jae Ryong; Kim, Jungwoo; Song, Chul-Hwa

2014-01-01

Highlights: • OECD/KAERI international CFD benchmark exercise was operated by KAERI. • The purpose is to validate relevant CFD codes based on the MATiS-H experiments. • Blind calculation results were synthesized in terms of mean velocity and RMS. • Quality of control volume rather than the number of it was emphasized. • Major findings were followed OECD/NEA CSNI report. - Abstract: The second international CFD benchmark exercise on turbulent mixing in a rod bundle has been launched by OECD/NEA, to validate relevant CFD (Computational Fluid Dynamics) codes and develop problem-specific Best Practice Guidelines (BPG) based on the KAERI (Korea Atomic Energy Research Institute) MATiS-H experiments on the turbulent mixing in a 5 × 5 rod array having two different types of vaned spacer grids: split and swirl types. For this 2nd international benchmark exercise (IBE-2), the MATiS-H testing provided a unique set of experimental data such as axial and lateral velocity components, turbulent intensity, and vorticity information. Blind CFD calculation results were submitted by twenty-five (25) participants to KAERI, who is the host organization of the IBE-2, and then analyzed and synthesized by comparing them with the MATiS-H data. Based on the synthesis of the results from both the experiments and blind CFD calculations for the IBE-2, and also by comparing with the IBE-1 benchmark exercise on the mixing in a T-junction, useful information for simulating this kind of complicated physical problem in a rod bundle was obtained. And some additional Best Practice Guidelines (BPG) are newly proposed. A summary of the synthesis results obtained in the IBE-2 is presented in this paper

Synthesis of the turbulent mixing in a rod bundle with vaned spacer grids based on the OECD-KAERI CFD benchmark exercise

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Ryong; Kim, Jungwoo; Song, Chul-Hwa, E-mail: chsong@kaeri.re.kr

2014-11-15

Highlights: • OECD/KAERI international CFD benchmark exercise was operated by KAERI. • The purpose is to validate relevant CFD codes based on the MATiS-H experiments. • Blind calculation results were synthesized in terms of mean velocity and RMS. • Quality of control volume rather than the number of it was emphasized. • Major findings were followed OECD/NEA CSNI report. - Abstract: The second international CFD benchmark exercise on turbulent mixing in a rod bundle has been launched by OECD/NEA, to validate relevant CFD (Computational Fluid Dynamics) codes and develop problem-specific Best Practice Guidelines (BPG) based on the KAERI (Korea Atomic Energy Research Institute) MATiS-H experiments on the turbulent mixing in a 5 × 5 rod array having two different types of vaned spacer grids: split and swirl types. For this 2nd international benchmark exercise (IBE-2), the MATiS-H testing provided a unique set of experimental data such as axial and lateral velocity components, turbulent intensity, and vorticity information. Blind CFD calculation results were submitted by twenty-five (25) participants to KAERI, who is the host organization of the IBE-2, and then analyzed and synthesized by comparing them with the MATiS-H data. Based on the synthesis of the results from both the experiments and blind CFD calculations for the IBE-2, and also by comparing with the IBE-1 benchmark exercise on the mixing in a T-junction, useful information for simulating this kind of complicated physical problem in a rod bundle was obtained. And some additional Best Practice Guidelines (BPG) are newly proposed. A summary of the synthesis results obtained in the IBE-2 is presented in this paper.

Design improvement for fretting-wear reduction of HANARO fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Cho, Yeong Garp; Chae, H. T.; Ryu, J. S.; Kim, H. R

2000-06-01

In the course of the visual inspection of the fuel assemblies un-loaded from the reactor core in December 1996, it was observed that many of fuel assemblies had mechanical damages on some components. The major damage was the freting-wear on spacer plates and endplates due to the flow induced vibration of the fuel assembly in the flow tube. Since the reactor is activated and the system modification for complete removal of the driving factors of the vibration of fuel assemblies is practically very difficult, the focus has been on the design change of the fuel assemblies. Consequently, various design changes were proposed to strengthen the wear resistance of the components based on the evaluation of the visual inspection results. The validity of the proposals was verified through the performance tests for the modified components, and the vibration test and endurance test for the fuel assemblies using the single-channel test rig(SCTR) in AECL.The subsequent design changes were additionally proposed based on the visual inspections for the fuel assemblies that had been fabricated according to the first design change and loaded in the core. As the effects of the first design change, the fretting-wear of spacer plates was remarkably reduced and the period until fretting-wear damage was extended by 60% for the first modified 36-rod fuel assembly. It is too early to say the endurance life time for the first modified 18-rod fuel assembly because of insufficient statistical data of only two bundles damaged, but the fretting-wear at the bottom endplate slot was reduced to about 50%. The second modified fuel assemblies, that were not loaded into the core yet, are expected to meet the design requirements for the core residence time due to strengthening the weak parts from the fretting-wear point of view. This report describes design changes and tests for fuel assemblies of HANARO to reduce the fretting-wear, and estimates the effects of design improvement quantitatively compared

Design improvement for fretting-wear reduction of HANARO fuel assembly

International Nuclear Information System (INIS)

Cho, Yeong Garp; Chae, H. T.; Ryu, J. S.; Kim, H. R.

2000-06-01

In the course of the visual inspection of the fuel assemblies un-loaded from the reactor core in December 1996, it was observed that many of fuel assemblies had mechanical damages on some components. The major damage was the freting-wear on spacer plates and endplates due to the flow induced vibration of the fuel assembly in the flow tube. Since the reactor is activated and the system modification for complete removal of the driving factors of the vibration of fuel assemblies is practically very difficult, the focus has been on the design change of the fuel assemblies. Consequently, various design changes were proposed to strengthen the wear resistance of the components based on the evaluation of the visual inspection results. The validity of the proposals was verified through the performance tests for the modified components, and the vibration test and endurance test for the fuel assemblies using the single-channel test rig(SCTR) in AECL.The subsequent design changes were additionally proposed based on the visual inspections for the fuel assemblies that had been fabricated according to the first design change and loaded in the core. As the effects of the first design change, the fretting-wear of spacer plates was remarkably reduced and the period until fretting-wear damage was extended by 60% for the first modified 36-rod fuel assembly. It is too early to say the endurance life time for the first modified 18-rod fuel assembly because of insufficient statistical data of only two bundles damaged, but the fretting-wear at the bottom endplate slot was reduced to about 50%. The second modified fuel assemblies, that were not loaded into the core yet, are expected to meet the design requirements for the core residence time due to strengthening the weak parts from the fretting-wear point of view. This report describes design changes and tests for fuel assemblies of HANARO to reduce the fretting-wear, and estimates the effects of design improvement quantitatively compared

Verification test of advanced LWR fuel components of Westinghouse type nuclear power plants

International Nuclear Information System (INIS)

Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Young Ho

2004-08-01

The purpose of this project is to independently conduct the performance test of the spacer grids and the cladding material of the 16x16 and 17x17 advanced fuels for Westinghouse type plants, and to improve the relevant test technology. Major works and results of the present research are as follows. 1. The design and structural features of the spacer grids were investigated, especially the finally determined I-spring was thoroughly analyzed in the point of the mechanical damage and characteristic. 2. As for the mechanical tests of the space grids, the characterization, the impact and the fretting wear tests were carried out. The block as well as the in-grid tests were conducted for the spring/dimple characterization, from which a simple method was developed that simulated the boundary conditions of the assembled grid straps. The impact tester was modified and improved to accommodate a full size grid assembly. The impact result showed that the grid assembly fulfilled the design criteria. As for the fretting wear tests, a sliding test under the room temperature air/water, a sliding/impact test under the room temperature air and a sliding/impact tests under the high temperature and pressure environments were carried out. To this end, a high temperature and pressure fretting wear tester was newly developed. The wear characteristic and the resistibility of the advanced grid spring/dimple were analyzed in detail. The test results were verified through comparing those with the test results by the Westinghouse company. 3. The properties and performance of the newly adopted material for the cladding, Low Sn Zirlo was investigated by a room and high temperature tensile tests and a corrosion tests under the environments of 360 .deg. C water, 400 steam and 360 .deg. C 70ppm LiOH. Through the present project, all the test equipment and technologies for the fuel components were procured, which will be used for future domestic development of a new fuel

CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

Energy Technology Data Exchange (ETDEWEB)

Pena, C.; Pellacani, F.; Macian Juan, R., E-mail: carlos.pena@ntech.mw.tum.de, E-mail: pellacani@ntech.mw.tum.de, E-mail: macian@ntech.mw.tum.de [Technische Universitaet Muenchen, Garching (Germany). Ntech Lehrstuhl fuer Nukleartechnik; Chiva, S., E-mail: schiva@emc.uji.es [Universitat Jaume I, Castellon de la Plana (Spain). Dept. de Ingenieria Mecanica y Construccion; Barrachina, T.; Miro, R., E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV) (Spain). Institute for Industrial, Radiophysical and Environmental Safety

2011-07-01

been developed for calculation and synchronization purposes. The data exchange is realized by means of the Parallel Virtual Machine (PVM) software package. In this contribution, steady-state and transient results of a quarter of PWR fuel assembly with cold water injection are presented and compared with obtained results from a RELAP5/PARCS v2.7 coupled calculation. A simplified model for the spacers has been included. A methodology has been introduced to take into account the pressure drop and the turbulence enhancement produced by the spacers. (author)

CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

International Nuclear Information System (INIS)

Pena, C.; Pellacani, F.; Macian Juan, R.; Chiva, S.; Barrachina, T.; Miro, R.

2011-01-01

developed for calculation and synchronization purposes. The data exchange is realized by means of the Parallel Virtual Machine (PVM) software package. In this contribution, steady-state and transient results of a quarter of PWR fuel assembly with cold water injection are presented and compared with obtained results from a RELAP5/PARCS v2.7 coupled calculation. A simplified model for the spacers has been included. A methodology has been introduced to take into account the pressure drop and the turbulence enhancement produced by the spacers. (author)

Hydrogel Spacer Prospective Multicenter Randomized Controlled Pivotal Trial: Dosimetric and Clinical Effects of Perirectal Spacer Application in Men Undergoing Prostate Image Guided Intensity Modulated Radiation Therapy

Energy Technology Data Exchange (ETDEWEB)

Mariados, Neil, E-mail: nmariados@ampofny.com [Associated Medical Professionals of New York, Syracuse, New York (United States); Sylvester, John [21st Century Oncology, East Bradenton, Florida (United States); Shah, Dhiren [Western New York Urology Associates, Cancer Care of WNY, Cheektowaga, New York (United States); Karsh, Lawrence [The Urology Center of Colorado, Denver, Colorado (United States); Hudes, Richard [Chesapeake Urology Research Associates, The Prostate Center, Owings Mills, Maryland (United States); Beyer, David [Arizona Oncology Services Foundation, Phoenix, Arizona (United States); Kurtzman, Steven [Urological Surgeons of Northern California, Campbell, California (United States); Bogart, Jeffrey [The Research Foundation of State University of New York, SUNY Upstate Medical University, Syracuse, New York (United States); Hsi, R. Alex [Peninsula Cancer Center, Poulsbo, Washington (United States); Kos, Michael [Urology Nevada, Reno, Nevada (United States); Ellis, Rodney [University Hospitals Case Medical Center, Cleveland, Ohio (United States); Logsdon, Mark [Sutter Health Sacramento Sierra Region, Sutter Institute for Medical Research, Sacramento, California (United States); Zimberg, Shawn [Advanced Radiation Centers of New York, Lake Success, New York (United States); Forsythe, Kevin [Oregon Urology Institute, Springfield, Oregon (United States); Zhang, Hong [University of Rochester, Rochester, New York (United States); Soffen, Edward [CentraState Medical Center, Freehold, New Jersey (United States); Francke, Patrick [Carolina Regional Cancer Center, 21st Century Oncology, Myrtle Beach, South Carolina (United States); Mantz, Constantine [21st Century Oncology, Fort Meyers, Florida (United States); Rossi, Peter [Emory University, Atlanta, Georgia (United States); DeWeese, Theodore [The Johns Hopkins University, Baltimore, Maryland (United States); and others

2015-08-01

Purpose: Perirectal spacing, whereby biomaterials are placed between the prostate and rectum, shows promise in reducing rectal dose during prostate cancer radiation therapy. A prospective multicenter randomized controlled pivotal trial was performed to assess outcomes following absorbable spacer (SpaceOAR system) implantation. Methods and Materials: Overall, 222 patients with clinical stage T1 or T2 prostate cancer underwent computed tomography (CT) and magnetic resonance imaging (MRI) scans for treatment planning, followed with fiducial marker placement, and were randomized to receive spacer injection or no injection (control). Patients received postprocedure CT and MRI planning scans and underwent image guided intensity modulated radiation therapy (79.2 Gy in 1.8-Gy fractions). Spacer safety and impact on rectal irradiation, toxicity, and quality of life were assessed throughout 15 months. Results: Spacer application was rated as “easy” or “very easy” 98.7% of the time, with a 99% hydrogel placement success rate. Perirectal spaces were 12.6 ± 3.9 mm and 1.6 ± 2.0 mm in the spacer and control groups, respectively. There were no device-related adverse events, rectal perforations, serious bleeding, or infections within either group. Pre-to postspacer plans had a significant reduction in mean rectal V70 (12.4% to 3.3%, P<.0001). Overall acute rectal adverse event rates were similar between groups, with fewer spacer patients experiencing rectal pain (P=.02). A significant reduction in late (3-15 months) rectal toxicity severity in the spacer group was observed (P=.04), with a 2.0% and 7.0% late rectal toxicity incidence in the spacer and control groups, respectively. There was no late rectal toxicity greater than grade 1 in the spacer group. At 15 months 11.6% and 21.4% of spacer and control patients, respectively, experienced 10-point declines in bowel quality of life. MRI scans at 12 months verified spacer absorption. Conclusions: Spacer

Grid Cell Relaxation Effects on the High Frequency Vibration Characteristics

International Nuclear Information System (INIS)

Ryu, Joo-Young; Eom, Kyong-Bo; Jeon, Sang-Youn; Kim, Jae-Ik

2015-01-01

The plate structure of the grid of fuel assembly is always exposed to serious vortex induced vibration. Also, High Frequency flow induced Vibration (HFV) is primarily generated by vortex-shedding effect. When it comes to grid design as a fuel assembly component, HFV should be considered in advance since it is one of the critical factors. Excessive HFV has a possibility of making degradation of the fuel reliability that is directly related to the fuel robustness and operating performance. KEPCO NF (KNF) has performed HFV tests with various grid designs. While studying the HFV characteristics through the HFV tests, it has been observed that HFV amplitudes show different levels according to grid cell relaxation. It means that the testing could give different interpretations due to the condition of grid cell. Since the amount of relaxation is different under operating conditions and environments in a reactor, test specimens should be modified as much as possible to the real state of the fuel. Therefore, in order to consider the grid cell relaxation effects on the HFV tests, it is important to use cell sized or non-cell sized grids. The main focus of this study is to find out how the HFV characteristics such as amplitude and frequency are affected by grid cell relaxation. Three cases of the grid cell sized specimen which is nickel alloy were prepared and tested. Through the comparison of the test results, it could be concluded that HFV amplitudes show decreasing trend according to the grid cell relaxation in the case of nickel alloy grid. It is also possible to expect the tendency of grid cell relaxation of a zirconium alloy grid based on test results

Post-Irradiation Examination Test of the Parts of X-Gen Nuclear Fuel Assembly

International Nuclear Information System (INIS)

Ahn, S. B.; Ryu, W. S.; Choo, Y. S.

2008-08-01

The mechanical properties of the parts of a nuclear fuel assembly are degraded during the operation of the reactor, through the mechanism of irradiation damage. The properties changes of the parts of the fuel assembly should be quantitatively estimated to ensure the safety of the fuel assembly and rod during the operation. The test techniques developed in this report are used to produce the irradiation data of the grid 1x1 cell spring, the grid 1x1 cell, the spring on one face of the 1x1 cell, the inner/outer strip of the grid and the welded part. The specimens were irradiated in the CT test hole of HANARO of a 30 MW thermal output at 300 deg. C during about 100 days From the spring test of mid grid 1x1 cell and grid plate, the irradiation effects can be examined. The irradiation effects on the irradiation growth also were occurred. The buckling load of mid grid 1x1 cell does not change with a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. The tensile test and microstructure examination of the spot and fillet welded parts are performed for the evaluation of an irradiation effects. Through these tests of components, the essential data on the fuel assembly design could be obtained. These results will be used to update the irradiation behavior databases, to improve the performance of fuel assembly, and to predict the service life of the fuel assembly in a reactor

Biofouling Control in Spiral-Wound Membrane Systems: Impact of Feed Spacer Modification and Biocides

KAUST Repository

Siddiqui, Amber

2016-12-01

High-quality drinking water can be produced with membrane-based filtration processes like reverse osmosis and nanofiltration. One of the major problems in these membrane systems is biofouling that reduces the membrane performance, increasing operational costs. Current biofouling control strategies such as pre-treatment, membrane modification, and chemical cleaning are not sufficient in all cases. Feed spacers are thin (0.8 mm), complex geometry meshes that separate membranes in a module. The main objective of this research was to evaluate whether feed spacer modification is a suitable strategy to control biofouling. Membrane fouling simulator studies with six feed spacers showed differences in biofouled spacer performance, concluding that (i) spacer geometry influences biofouling impact and (ii) biofouling studies are essential for evaluation of spacer biofouling impact. Computed tomography (CT) was found as a suitable technique to obtain three-dimensional (3D) measurements of spacers, enabling more representative mathematical modeling of hydraulic behavior of spacers in membrane systems. A strategy for developing, characterizing, and testing of spacers by numerical modeling, 3D printing of spacers and experimental membrane fouling simulator studies was developed. The combination of modeling and experimental testing of 3D printed spacers is a promising strategy to develop advanced spacers aiming to reduce the impact of biofilm formation on membrane performance and to improve the cleanability of spiral-wound membrane systems.

Ertapenem Articulating Spacer for the Treatment of Polymicrobial Total Knee Arthroplasty Infection

Directory of Open Access Journals (Sweden)

Dragan Radoicic

2016-01-01

Full Text Available Introduction. Periprosthetic joint infections (PJIs are the primary cause of early failure of the total knee arthroplasty (TKA. Polymicrobial TKA infections are often associated with a higher risk of treatment failure. The aim of the study was to assess the efficacy of ertapenem loaded spacers in the treatment of polymicrobial PJI. Methods. There were 18 patients enrolled; nine patients with polymicrobial PJI treated with ertapenem loaded articulating spacers were compared to the group of 9 patients treated with vancomycin or ceftazidime loaded spacers. Results. Successful reimplantation with revision implants was possible in 66.67%. Ertapenem spacers were used in 6 cases in primary two-stage procedure and in 3 cases in secondary spacer exchange. Successful infection eradication was achieved in all cases; final reimplantation with revision knee arthroplasty implants was possible in 6 cases. Conclusion. Ertapenem can be successfully used as antimicrobial addition to the cement spacers in two-stage revision treatment of polymicrobial PJIs. However, this type of spacer may also be useful in the treatment of infections caused by monomicrobial extended spectrum beta-lactamases producing gram-negative bacilli. Further clinical studies are required to evaluate the efficacy and safety of ertapenem spacers in the treatment of polymicrobial and monomicrobial PJIs.

Study and modeling of fluctuating fluid forces exerted on fuel rods in pressurized water reactors

International Nuclear Information System (INIS)

Bhattacharjee, Saptarshi

2016-01-01

Flow-induced vibrations in a pressurized water reactor (PWR) core can cause fretting wear in the fuel rods. Due to friction, wear occurs at the contact locations between the spacer grid and the fuel rod. This could compromise the first safety barrier of the nuclear reactor by damaging the fuel rod cladding. In order to ensure the integrity of the cladding, it is necessary to know the random fluctuating forces acting on the rods. However, the spectra for these fluid forces are not well known. The goal of this PhD thesis was to use simple geometrical elements to check the reproducibility of realistic pressurized water reactor spacer grids. As a first step, large eddy simulations were performed on a concentric annular pipe for different mesh refinements using the CFD code Trio CFD (previously Trio U) developed by CEA. A mesh sensitivity study was performed to obtain an acceptable mesh for reproducing standard literature results. This information on mesh resolution was used when carrying out simulations using various geometric obstacles inside the pipe, namely, mixing vanes, circular spacer grid and a combination of square spacer grid with mixing vanes. The last of the three configurations is the closest to a realistic PWR fuel assembly. Structured mesh was generated for the annular pipe case and circular grid case. An innovative hybrid mesh was used for the two remaining cases of the mixing vanes and the square grid: keeping unstructured mesh around the obstacles and structured mesh in the rest of the domain. The inner wall of the domain was representative of the fuel rod cladding. Both hydraulic and wall pressure characteristics were analyzed for each case. The results for the square grid case were found to be an approximate combination of the mixing vane case and circular grid case. Simulation results were compared with experiments performed at CEA Cadarache. Some preliminary comparisons were also made with classical semi-empirical models. (author) [fr

Grids for nuclear fuel elements

International Nuclear Information System (INIS)

Nicholson, G.

1980-01-01

This invention relates to grids for nuclear fuel assemblies with the object of providing an improved grid, tending to have greater strength and tending to offer better location of the fuel pins. It comprises sets of generally parallel strips arranged to intersect to define a structure of cellular form, at least some of the intersections including a strip which is keyed to another strip at more than one point. One type of strip may be dimpled along its length and another type of strip may have slots for keying with the dimples. (Auth.)

Computer-Aided Design Method of Warp-Knitted Jacquard Spacer Fabrics

Directory of Open Access Journals (Sweden)

Li Xinxin

2016-06-01

Full Text Available Based on a further study on knitting and jacquard principles, this paper presents a mathematical design model to make computer-aided design of warp-knitted jacquard spacer fabrics more efficient. The mathematical model with matrix method employs three essential elements of chain notation, threading and Jacquard designing. With this model, the processing to design warp-knitted jacquard spacer fabrics with CAD software is also introduced. In this study, the sports shoes which have separated functional areas according to the feet structure and characteristics of movement are analysed. The results show the different patterns on Jacquard spacer fabrics that are seamlessly stitched with jacquard technics. The computer-aided design method of warp-knitted jacquard spacer fabrics is efficient and simple.

Evolution of fuel rod support under irradiation impact on the mechanical behaviour of fuel assemblies

International Nuclear Information System (INIS)

Billerey, Antoine; Waeckel, Nicolas

2005-01-01

New fuel management targets imply to increase fuel assembly discharge burnup. Therefore, the prediction of the mechanical behaviour of the irradiated fuel assembly is essential such as excessive fuel assembly distortion induce incomplete Rod Cluster Control Assembly insertion problems (safety issue) or fuel rod vibration induced wear leading to leaking rods (plant operation problems). Within this framework, one of the most important parameter is the knowledge of the fuel rod support in the grid cell because it directly governs the mechanical behaviour of the fuel assembly and consequently allows to predict the behaviour of irradiated structures in terms of (1) axial and lateral deformation (global behaviour of the assembly) and (2) rod vibration induced wear (local behaviour of the rod). Generally, fuel rod support is provided by a spring-dimple system fixed to the grid. During irradiation, the spring force decreases and a gap between the rod and the spring may occur. This phenomenon is due to (1) stress relieving in the spring and in the dimples, (2) grid growth and (3) reduction of the rod diameter. Two models have been developed to predict the behaviour of the rod in the cell. The first model is dedicated to the evaluation of the spring force relaxation during irradiation. The second one can assess the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (1) the creep laws of the grid materials, (2) the growth law of the grid, (3) the evolution of rod diameter and (4) the design of the fuel rod support. The aim of this paper is to: (1) evaluate the consequences of grid support design modifications on the rod vibration sensitivity in terms of predicted rod to grid maximum gap during irradiation and time in operation with an open rod to grid gap, (2) evaluate, using a linear or non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the overall mechanical

Nuclear reactor fuel assembly

International Nuclear Information System (INIS)

1975-01-01

A description is given of a nuclear reactor fuel assembly comprising a cluster of fuel elements supported by transversal grids so that their axes are parallel to and at a distance from each other, in order to establish interstices for the axial flow of a coolant. At least one of the interstices is occupied by an axial duct reserved for an auxiliary cooling fluid and is fitted with side holes through which the auxiliary cooling fluid is sprayed into the cluster. Deflectors extend as from a transversal grid in a position opposite the holes to deflect the cooling fluid jet towards those parts of the fuel elements that are not accessible to the auxiliary coolant. This assembly is intended for reactors cooled by light or heavy water [fr

Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly. Final CRADA Report

International Nuclear Information System (INIS)

Tentner, A.

2009-01-01

A direct numerical simulation capability for two-phase flows with heat transfer in complex geometries can considerably reduce the hardware development cycle, facilitate the optimization and reduce the costs of testing of various industrial facilities, such as nuclear power plants, steam generators, steam condensers, liquid cooling systems, heat exchangers, distillers, and boilers. Specifically, the phenomena occurring in a two-phase coolant flow in a BWR (Boiling Water Reactor) fuel assembly include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for this purpose of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Advanced CFD (Computational Fluid Dynamics) codes provide a potential for detailed 3D simulations of coolant flow inside a fuel assembly, including flow around a spacer element using more fundamental physical models of flow regimes and phase interactions than sub-channel codes. Such models can extend the code applicability to a wider range of situations, which is highly important for increasing the efficiency and to prevent accidents.

Fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Ferrari, H.M.; Miller, D.L.; Tong, L.S.

1975-01-01

A description is given of a fuel assembly including multiple open channel grids for holding fuel rods and control rod guide thimbles in predetermined fixed relationship with each other. Metallic straps are interwoven to form a grid or egg crate configuration having openings which receive the fuel rods and guide thimbles. To properly support and cool the fuel rods near the grid-fuel rod interface, springs and dimples on the grid straps project into each opening, the dimples being oriented in a direction to permit flow of coolant upwardly therethrough. To minimize turbulence in coolant flow, the leading edge of each grid strap is provided with cutout sections which form scallops effective in channeling coolant in a uniform flow path through the network of grid openings

Replacement rod

International Nuclear Information System (INIS)

Hatfield, S.C.

1989-01-01

This patent describes in an elongated replacement rod for use with fuel assemblies of the type having two end fittings connected by guide tubes with a plurality of rod and guide tube cell defining spacer grids containing rod support features and mixing vanes. The grids secured to the guide tubes in register between the end fittings at spaced intervals. The fuel rod comprising: an asymmetrically beveled tip; a shank portion having a straight centerline; and a permanently diverging portion between the tip and the shank portion

Impact of spacer thickness on biofouling in forward osmosis

KAUST Repository

Valladares Linares, Rodrigo

2014-06-01

Forward osmosis (FO) indirect desalination systems integrate wastewater recovery with seawater desalination. Niche applications for FO systems have been reported recently, due to the demonstrated advantages compared to conventional high-pressure membrane processes such as nanofiltration (NF) and reverse osmosis (RO). Among them, wastewater recovery has been identified to be particularly suitable for practical applications. However, biofouling in FO membranes has rarely been studied in applications involving wastewater effluents. Feed spacers separating the membrane sheets in cross-flow systems play an important role in biofilm formation. The objective of this study was to determine the influence of feed spacer thickness (28, 31 and 46mil) on biofouling development and membrane performance in a FO system, using identical cross-flow cells in parallel studies. Flux development, biomass accumulation, fouling localization and composition were determined and analyzed. For all spacer thicknesses, operated at the same feed flow and the same run time, the same amount of biomass was found, while the flux reduction decreased with thicker spacers. These observations are in good agreement with biofouling studies for RO systems, considering the key differences between FO and RO. Our findings contradict previous cross-flow studies on particulate/colloidal fouling, where higher cross-flow velocities improved system performance. Thicker spacers reduced the impact of biofouling on FO membrane flux. © 2014 Elsevier Ltd.

Impact of spacer thickness on biofouling in forward osmosis.

Science.gov (United States)

Valladares Linares, R; Bucs, Sz S; Li, Z; AbuGhdeeb, M; Amy, G; Vrouwenvelder, J S

2014-06-15

Forward osmosis (FO) indirect desalination systems integrate wastewater recovery with seawater desalination. Niche applications for FO systems have been reported recently, due to the demonstrated advantages compared to conventional high-pressure membrane processes such as nanofiltration (NF) and reverse osmosis (RO). Among them, wastewater recovery has been identified to be particularly suitable for practical applications. However, biofouling in FO membranes has rarely been studied in applications involving wastewater effluents. Feed spacers separating the membrane sheets in cross-flow systems play an important role in biofilm formation. The objective of this study was to determine the influence of feed spacer thickness (28, 31 and 46 mil) on biofouling development and membrane performance in a FO system, using identical cross-flow cells in parallel studies. Flux development, biomass accumulation, fouling localization and composition were determined and analyzed. For all spacer thicknesses, operated at the same feed flow and the same run time, the same amount of biomass was found, while the flux reduction decreased with thicker spacers. These observations are in good agreement with biofouling studies for RO systems, considering the key differences between FO and RO. Our findings contradict previous cross-flow studies on particulate/colloidal fouling, where higher cross-flow velocities improved system performance. Thicker spacers reduced the impact of biofouling on FO membrane flux. Copyright © 2014 Elsevier Ltd. All rights reserved.

Impact of spacer thickness on biofouling in forward osmosis

KAUST Repository

Valladares Linares, Rodrigo; Bucs, Szilard; Li, Z.; AbuGhdeeb, M.; Amy, Gary L.; Vrouwenvelder, Johannes S.

2014-01-01

Forward osmosis (FO) indirect desalination systems integrate wastewater recovery with seawater desalination. Niche applications for FO systems have been reported recently, due to the demonstrated advantages compared to conventional high-pressure membrane processes such as nanofiltration (NF) and reverse osmosis (RO). Among them, wastewater recovery has been identified to be particularly suitable for practical applications. However, biofouling in FO membranes has rarely been studied in applications involving wastewater effluents. Feed spacers separating the membrane sheets in cross-flow systems play an important role in biofilm formation. The objective of this study was to determine the influence of feed spacer thickness (28, 31 and 46mil) on biofouling development and membrane performance in a FO system, using identical cross-flow cells in parallel studies. Flux development, biomass accumulation, fouling localization and composition were determined and analyzed. For all spacer thicknesses, operated at the same feed flow and the same run time, the same amount of biomass was found, while the flux reduction decreased with thicker spacers. These observations are in good agreement with biofouling studies for RO systems, considering the key differences between FO and RO. Our findings contradict previous cross-flow studies on particulate/colloidal fouling, where higher cross-flow velocities improved system performance. Thicker spacers reduced the impact of biofouling on FO membrane flux. © 2014 Elsevier Ltd.

Evolution of fuel rod support under irradiation consequences on the mechanical behavior of fuel assembly

International Nuclear Information System (INIS)

Billerey, A.; Bouffioux, P.

2002-01-01

The complete paper follows. According to the fuel management policy in French PWR with respect to high burn-up, the prediction of the mechanical behavior of the irradiated fuel assembly is required as far as excessive deformations of fuel assembly might lead to incomplete Rod Cluster Control Assembly insertion (safety problems) and fretting wear lead to leaking rods (plant operation problems). One of the most important parameter is the evolution of the fuel rod support in the grid cell as it directly governs the mechanical behavior of the fuel assembly and consequently allows to predict the behavior of irradiated structure in terms of (i) axial and lateral deformation (global behavior of the assembly) and (ii) fretting wear (local behavior of the rod). Fuel rod support is provided by a spring-dimple system fixed on the grid. During irradiation, the spring force decreases and a gap between the rod and the spring might open. This phenomenon is due to (i) irradiation-induced stress relaxation for the spring and for the dimples, (ii) grid growth and (iii) reduction of rod diameter. Two models have been developed to predict the behavior of the rod in the grid cell. The first model is able to evaluate the spring force relaxation during irradiation. The second one is able to evaluate the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (i) the creep laws of the grid materials, (ii) the growth law of the grid, (iii) the evolution of rod diameter and (iv) the design of the fuel rod support. The objectives of this paper are to: (i) evaluate the consequences of grid support design modifications on the fretting sensitivity in terms of predicted maximum gap during irradiation and operational time to gap appearance; (ii) evaluate, using a non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the mechanical behavior of the full assembly in terms of axial and

Optimized Design of Spacer in Electrodialyzer Using CFD Simulation Method

Science.gov (United States)

Jia, Yuxiang; Yan, Chunsheng; Chen, Lijun; Hu, Yangdong

2018-06-01

In this study, the effects of length-width ratio and diversion trench of the spacer on the fluid flow behavior in an electrodialyzer have been investigated through CFD simulation method. The relevant information, including the pressure drop, velocity vector distribution and shear stress distribution, demonstrates the importance of optimized design of the spacer in an electrodialysis process. The results show width of the diversion trench has a great effect on the fluid flow compared with length. Increase of the diversion trench width could strength the fluid flow, but also increase the pressure drop. Secondly, the dead zone of the fluid flow decreases with increase of length-width ratio of the spacer, but the pressure drop increases with the increase of length-width ratio of the spacer. So the appropriate length-width ratio of the space should be moderate.

TVSA-T fuel assembly for 'Temelin' NPP. Main results of design and safety analyses. Trends of development

International Nuclear Information System (INIS)

Samojlov, O.B.; Kajdalov, V.B.; Falkov, A.A.; Bolnov, V.A.; Morozkin, O.N.; Molchanov, V.L.; Ugryumov, A.V.

2010-01-01

TVSA is a fuel assembly with rigid skeleton formed by 6 angle pieces and SG is successfully operated at 17 VVER-1000 power units of Kalinin NPP, as well as at Ukrainian and Bulgarian NPPs. Based on a contract for fuel supply to the Temelin NPP, the TVSA-T fuel assembly was developed, building on proven solutions confirmed by operation of TVSA modifications during 4-6 years and by the results of post-irradiation examination. The TVSA-T design includes combined spacer grids (SG+MG) and by fuel column elongation by 150 mm. A set of analyses and experiments was performed to validate the design, including thermal hydraulic tests, validation of critical heat flux correlation for TVSA-T, integrated mechanical, vibration and lifetime tests. A licence to use the fuel has been granted by the Czech State Office for Nuclear Safety. The TVSA-T core is currently in operation at the Temelin-1 reactor unit. The presentation is concluded as follows: TVSA-T fuel assembly for Temelin has been validated. The TVSA-T design is based on approved technical decisions and meets the current requirements for lifetime, operational maneuverability and safety. The results of post-irradiation examination of TVSA-T operated at the Kalinin-1 unit for 4 years confirm the assembly operability, skeleton stiffness, geometric stability and normal fuel rod cladding condition. The properties of the TVSA fuel with MG allow the core power to be increased up to 3300 MW to match the envisaged future VVER (MIR-1200) design, providing allowable fuel rod power FΔh =1.63 (to implement effective fuel cycles). (P.A.)

Heterogeneous diversity of spacers within CRISPR

Science.gov (United States)

Deem, Michael; He, Jiankui

2011-03-01

Clustered regularly interspaced short palindromic repeats (CRISPR) in bacterial and archaeal DNA have recently been shown to be a new type of anti-viral immune system in these organisms. We here study the diversity of spacers in CRISPR under selective pressure. We propose a population dynamics model that explains the biological observation that the leader-proximal end of CRISPR is more diversified and the leader-distal end of CRISPR is more conserved. This result is shown to be in agreement with recent experiments. Our results show that the CRISPR spacer structure is influenced by and provides a record of the viral challenges that bacteria face. 1) J. He and M. W. Deem, Phys. Rev. Lett. 105 (2010) 128102

Prospects for investment in large-scale, grid-connected solar power in Africa

DEFF Research Database (Denmark)

Hansen, Ulrich Elmer; Nygaard, Ivan; Pedersen, Mathilde Brix

since the 1990s have changed the competiveness of solar PV in all markets, ranging from individual households via institutions to mini-grids and grid-connected installations. In volume and investment, the market for large-scale grid-connected solar power plants is by far the most important......-scale investments in grid-connected solar power plants and local assembly facilities for PV panels, have exceeded even optimistic scenarios. Finally, therefore, there seem to be bright prospects for investment in large-scale grid-connected solar power in Africa....

ATLAS barrel hadron tile calorimeter: spacers plates mass production

International Nuclear Information System (INIS)

Artikov, A.M.; Budagov, Yu.A.; Khubua, J.

1999-01-01

In this article we expose the main problems of the mass production of the so-called 'spacer plates' for the ATLAS Barrel Hadron Tile Calorimeter. We describe all practical solutions of these problems. Particularly we present the measurement procedures and calculation schemes we used for the spacers dimensions determination. The results of the calculations are presented

Impact of feed spacer and membrane modification by hydrophilic, bactericidal and biocidal coating on biofouling control

KAUST Repository

Araújo, Paula A.

2012-06-01

The influence of polydopamine- and polydopamine-. graft-poly(ethylene glycol)-coated feed spacers and membranes, copper-coated feed spacers, and commercially-available biostatic feed spacers on biofouling has been studied in membrane fouling simulators. Feed spacers and membranes applied in practical membrane filtration systems were used; biofouling development was monitored by feed channel pressure drop increase and biomass accumulation. Polydopamine and polydopamine-. g-PEG are hydrophilic surface modification agents expected to resist protein and bacterial adhesion, while copper feed spacer coatings and biocides infused in feed spacers are expected to restrict biological growth. Our studies showed that polydopamine and polydopamine-. g-PEG coatings on feed spacers and membranes, copper coatings on feed spacers, and a commercial biostatic feed spacer did not have a significant impact on feed channel pressure drop increase and biofilm accumulation as measured by ATP and TOC content. The studied spacer and membrane modifications were not effective for biofouling control; it is doubtful that feed spacer and membrane modification, in general, may be effective for biofouling control regardless of the type of applied coating. © 2012 Elsevier B.V.

Transcriptional regulator-mediated activation of adaptation genes triggers CRISPR de novo spacer acquisition

DEFF Research Database (Denmark)

Liu, Tao; Li, Yingjun; Wang, Xiaodi

2015-01-01

Acquisition of de novo spacer sequences confers CRISPR-Cas with a memory to defend against invading genetic elements. However, the mechanism of regulation of CRISPR spacer acquisition remains unknown. Here we examine the transcriptional regulation of the conserved spacer acquisition genes in Type I......, it was demonstrated that the transcription level of csa1, cas1, cas2 and cas4 was significantly enhanced in a csa3a-overexpression strain and, moreover, the Csa1 and Cas1 protein levels were increased in this strain. Furthermore, we demonstrated the hyperactive uptake of unique spacers within both CRISPR loci...... in the presence of the csa3a overexpression vector. The spacer acquisition process is dependent on the CCN PAM sequence and protospacer selection is random and non-directional. These results suggested a regulation mechanism of CRISPR spacer acquisition where a single transcriptional regulator senses the presence...

Polyetheretherketone (PEEK) Spacers for Anterior Cervical Fusion: A Retrospective Comparative Effectiveness Clinical Trial.

Science.gov (United States)

Lemcke, Johannes; Al-Zain, Ferass; Meier, Ullrich; Suess, Olaf

2011-01-01

Anterior cervical decompression and fusion (ACDF) is the standard surgical treatment for radiculopathy and myelopathy. Polyetheretherketone (PEEK) has an elasticity similar to bone and thus appears well suited for use as the implant in ACDF procedures. The aim of this study is to examine the clinical and radiographic outcome of patients treated with standing alone PEEK spacers without bone morphogenic protein (BMP) or plating and to examine the influence of the different design of the two spacers on the rate of subsidence and dislocation. This retrospective comparative study reviewed 335 patients treated by ACDF in a specialized urban hospital for radiculopathy or myelopathy due to degenerative pathologies. The Intromed PEEK spacer was used in 181 patients from 3/2002 to 11/2004, and the AMT SHELL spacer was implanted in 154 patients from 4/2004 to 12/2007. The follow-up rate was 100% at three months post-op and 82.7% (277 patients) at one year. The patients were assessed with the Japanese Orthopedic Association (JOA) questionnaire and radiographically. At the one-year follow-up there were 118/277 patients with an excellent clinical outcome on the JOA, 112/277 with a good outcome, 20/277 with a fair outcome, and 27/277 with a poor outcome. Subsidence was observed in 13.3% of patients with the Intromed spacer vs 8.4% of the patients with the AMT SHELL. Dislocation of the spacer was observed in 10 of the 181 patients with Intromed spacers but in none of the 154 patients with Shell spacers. The study demonstrates that ACDF with standing alone PEEK cages leads to excellent and good clinical outcomes. The differences we observed in the subsidence rate between the two spacers were not significant and cannot be related to a single design feature of the spacers.

Economic analysis of grid and wire wrap supported hydride and oxide fueled pressurized water reactors

International Nuclear Information System (INIS)

Shuffler, C.; Diller, P.; Malen, J.; Todreas, N.; Greenspan, E.; Petrovic, B.

2009-01-01

An economic analysis is performed to calculate the levelized unit cost of electricity (COE) for a pressurized water reactor (PWR) retrofitted with a range of potential U (45 wt.%)-ZrH 1.6 hydride and UO 2 oxide fueled geometries (i.e., combinations of rod diameter and pitch) supported by traditional grid spacers (square array) and wire wrap spacers (hexagonal array). The time frame considered in computing the COE is the remaining plant life, beginning at the time of retrofit. The goals of the analysis are twofold: (1) comparing the economic performance of UO 2 and U-ZrH 1.6 fuels for a range of retrofitted geometries supported by grid and wire wrap spacers; and (2) investigating the potential economic benefits for nuclear utilities considering retrofitting new fuels and/or geometries into existing PWR pressure vessels. Fuel cycle, operations and maintenance (O and M), and capital costs are considered. The economic performance of U-ZrH 1.6 and UO 2 fuels is found to be similar, with UO 2 fueled designs providing a slight advantage when supported by grid spacers, and U-ZrH 1.6 providing a slight advantage when supported by wire wrap spacers. These small differences in cost, however, are within the bounds of uncertainty of this study and are not believed to provide a strong economic argument for the use of one fuel type over the other. To demonstrate the potential economic benefits of retrofitted designs to nuclear utilities, two different comparisons are made. The first compares the COE for retrofitted designs with the COE for a reference PWR, assumed to have operated long enough to recuperate its initial capital investment. The costs for this reference PWR reflect the 'do-nothing' case for current plant owners whose primary expenditures are fuel cycle and O and M costs. The second comparison introduces a different reference PWR that includes the costs to operate an existing unit and the cost to purchase power from a newly constructed PWR, for comparison with

Success of Two-Stage Reimplantation in Patients Requiring an Interim Spacer Exchange.

Science.gov (United States)

George, Jaiben; Miller, Evan M; Curtis, Gannon L; Klika, Alison K; Barsoum, Wael K; Mont, Michael A; Higuera, Carlos A

2018-03-23

Some patients undergoing a 2-stage revision for a periprosthetic joint infection require a repeat spacer in the interim (removal of existing spacer with insertion of a new spacer or spacer exchange) due to persistent infection. The objectives of this study are to (1) determine the factors associated with patients who receive a repeat spacer and (2) compare the infection-free survival (overall and stratified by joint type) of reimplantation in patients who did or did not receive a repeat spacer. From 2001 to 2014, 347 hip or knee 2-stage revisions that finally underwent reimplantation and had a minimum 2-year follow-up were identified. An interim spacer exchange was performed in 59 (17%) patients (exchange cohort). Patient-related and organism-related factors were compared between the exchange and non-exchange cohorts. Kaplan-Meier survival curves were performed to assess the success (absence of signs of infection, reoperation for infection, periprosthetic joint infection-related mortality) of both cohorts. Patients in the exchange group had higher comorbidity score (P = .020), prolonged time to reimplantation (P exchange cohort, and 78% (knee 77%, hip 78%) in the non-exchange cohort (P = .020). Patients requiring an interim spacer exchange were found to have more comorbidities, prolonged treatment period, and were more likely to be infected with a resistant organism. About one-third of such patients became reinfected within 5 years compared to only one-fifth of the patients without an interim spacer exchange. Copyright © 2018 Elsevier Inc. All rights reserved.

Separator-spacer for electrochemical systems

Science.gov (United States)

Grimes, Patrick G.; Einstein, Harry; Newby, Kenneth R.; Bellows, Richard J.

1983-08-02

An electrochemical cell construction features a novel co-extruded plastic electrode in an interleaved construction with a novel integral separator-spacer. Also featured is a leak and impact resistant construction for preventing the spill of corrosive materials in the event of rupture.

Effects of different rod spacers (helical types) on coolant crossmixing

International Nuclear Information System (INIS)

Zhukov, A.V.; Sviridenko, E.Ya.; Matyukhin, N.M.; Rymkevich, K.S.; Ushakov, P.A.

1981-11-01

The results of investigations (electromagnetic measuring method) on coolant cross mixing in rod clusters with spiral wire spacers with different winding directions, with alternating unfinned and finned rods (case 'fin to rod'), as well as in rod clusters with much space between the rods, (case 'fin to fin') are reported. The local fluid dynamics parameters (distribution of the transversal and longitudinal velocity component) that define the physical processes of the coolant exchange in the rod clusters with helical spacers are explained. The investigation results for different helical spacer types are compared with each other. (orig.) [de

VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

International Nuclear Information System (INIS)

Doshi, P.K.; DeMario, E.E.; Knott, R.P.

1993-01-01

Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR's) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design

Solution-processed organic tandem solar cells with embedded optical spacers

NARCIS (Netherlands)

Hadipour, Afshin; de Boer, Bert; Blom, Paul W. M.

2007-01-01

We demonstrate a solution-processed polymer tandem solar cell in which the two photoactive single cells are separated by an optical spacer. The use of an optical spacer allows for an independent optimization of both the electronic and optical properties of the tandem cell. The optical transmission

Predictable self-assembled [2×2] Ln(III)4 square grids (Ln = Dy,Tb)-SMM behaviour in a new lanthanide cluster motif.

Science.gov (United States)

Anwar, Muhammad Usman; Thompson, Laurence Kenneth; Dawe, Louise Nicole; Habib, Fatemah; Murugesu, Muralee

2012-05-14

The ditopic carbohydrazone ligand (L1) produces the square, self-assembled [2×2] grids [Dy(4)(L1)(4)(OH)(4)]Cl(2) (1) and [Ln(4)(L1)(4)(μ(4)-O)(μ(2)-1,1-N(3))(4)] (Ln = Dy (2), Tb (3)), with 2 exhibiting SMM behaviour. Two relaxation processes occur with U(eff) = 51 K, 91 K in the absence of an external field, and one with U(eff) = 270 K in the presence of a 1600 Oe optimum field. This journal is © The Royal Society of Chemistry 2012

Reflooding experiments on a 49-rod cluster containing a long 90% blockage

International Nuclear Information System (INIS)

Pearson, K.G.; Cooper, C.A.; Jowitt, D.; Kinneir, J.H.

1983-01-01

A series of reflooding experiments was performed on a model fuel assembly, containing a very severe partial blockage, in the THETIS rig. The assembly comprised 49 full length, electrically heated fuel rod simulators and the blockage was created by attaching thin-walled, preformed swellings to a group of 16 rods. Results are presented for single phase and forced reflooding experiments. The most important findings relate to the improvements in heat transfer created by spacer grids and the nature of the heat transfer processes within the blockage. Spacer grids are shown to improve heat transfer by increasing turbulence and also, when wet, by cooling the steam flowing through them. Liquid penetration evidently deteriorates as the rewetting front approaches the blockage, allowing the steam through the blockage to superheat strongly and giving rise to a late peak in cladding temperature. At low reflooding rates there is a temperature penalty associated with the blockage which becomes increasingly larger as the reflooding rate is reduced. The adequacy of cooling in this very severe blockage becomes questionable when the reflooding rate falls to about 2cm/s. (U.K.)

Spacer geometry and particle deposition in spiral wound membrane feed channels

KAUST Repository

Radu, A.I.

2014-11-01

Deposition of microspheres mimicking bacterial cells was studied experimentally and with a numerical model in feed spacer membrane channels, as used in spiral wound nanofiltration (NF) and reverse osmosis (RO) membrane systems. In-situ microscopic observations in membrane fouling simulators revealed formation of specific particle deposition patterns for different diamond and ladder feed spacer orientations. A three-dimensional numerical model combining fluid flow with a Lagrangian approach for particle trajectory calculations could describe very well the in-situ observations on particle deposition in flow cells. Feed spacer geometry, positioning and cross-flow velocity sensitively influenced the particle transport and deposition patterns. The deposition patterns were not influenced by permeate production. This combined experimental-modeling approach could be used for feed spacer geometry optimization studies for reduced (bio)fouling. © 2014 Elsevier Ltd.

Study on effects of mixing vane grids on coolant temperature distribution by subchannel analysis

Energy Technology Data Exchange (ETDEWEB)

Mao, H.; Yang, B.W.; Han, B. [Xi' an Jiaotong Univ., Shaanxi (China). Science and Technology Center for Advanced Nuclear Fuel Research

2016-07-15

Mixing vane grids (MVG) have great influence on coolant temperature field in the rod bundle. The MVG could enhance convective heat transfer between the fuel rod wall and the coolant, and promote inter-subchannel mixing at the same time. For the influence of the MVG on convective heat transfer enhancement, many experiments have been done and several correlations have been developed based on the experimental data. However, inter-subchannel mixing promotion caused by the MVG is not well estimated in subchannel analysis because the information of mixing vanes is totally missing in most subchannel codes. This paper analyzes the influence of mixing vanes on coolant temperature distribution using the improved MVG model in subchannel analysis. The coolant temperature distributions with the MVG are analyzed, and the results show that mixing vanes lead to a more uniform temperature distribution. The performances of split vane grids under different power conditions are evaluated. The results are compared with those of spacer grids without mixing vanes and some conclusions are obtained.

Study on effects of mixing vane grids on coolant temperature distribution by subchannel analysis

International Nuclear Information System (INIS)

Mao, H.; Yang, B.W.; Han, B.

2016-01-01

Mixing vane grids (MVG) have great influence on coolant temperature field in the rod bundle. The MVG could enhance convective heat transfer between the fuel rod wall and the coolant, and promote inter-subchannel mixing at the same time. For the influence of the MVG on convective heat transfer enhancement, many experiments have been done and several correlations have been developed based on the experimental data. However, inter-subchannel mixing promotion caused by the MVG is not well estimated in subchannel analysis because the information of mixing vanes is totally missing in most subchannel codes. This paper analyzes the influence of mixing vanes on coolant temperature distribution using the improved MVG model in subchannel analysis. The coolant temperature distributions with the MVG are analyzed, and the results show that mixing vanes lead to a more uniform temperature distribution. The performances of split vane grids under different power conditions are evaluated. The results are compared with those of spacer grids without mixing vanes and some conclusions are obtained.

ZZ PWR-AXBUPRO-GKN, Measured Axial Burnup Profiles, NPP Neckarewstheim

International Nuclear Information System (INIS)

Neuber, Jens-Christian; Lamprecht, Thomas

1999-01-01

Description or function: PWR-AXBUPRO-GKN12 contains Axial Burnup Shapes released by Siemens AG Power Generation Group. It contains data sets relative to following NPP and initial enrichment: - NPP Neckarwestheim 1, Fuel assemblies with an initial enrichment of 3.5 wt.-% 235-U; - NPP Neckarwestheim 2, Fuel assemblies with an initial enrichment of 3.5 wt.-% 235-U; - NPP Neckarwestheim 2, Fuel assemblies with an initial enrichment of 3.8 wt.-% 235-U; - NPP Neckarwestheim 2, Fuel assemblies with an initial enrichment of 4.0 wt.-% 235-U. In each of these files the axial shapes are listed one after the other. Each shape is characterised by: - the number of the cycle; - the number of the fuel assembly. - axial burnup shape characteristics: - height (in centimeter) of the nodes with respect to the active length of the fuel assemblies normalized to the cold, unirradiated state. - the nodal burnup (in MWd/kg U). - fuel assembly design data as well as core geometry and operating data pertinent to depletion calculations: NPP Neckarwestheim 1 (GKN1) - (square pitch lattice 15X15 - thermal Power 2497 MW)thermal Power 2497 MW); More than 700 EOC axial shapes from cycle 18 up. From cycle 18 to cycle 20 a change from an Out-In-Loading to an In-Out-Loading has taken place. Fuel assemblies up to number 1093 have spacer grids made of Inconel, whereas all the fuel assemblies from number 1094 up have spacer grids made of Zircaloy. Discharge burnups range from: 9.7 to 52.8 MWd/kg. NPP Neckarwestheim 2 (GKN2) - (square pitch lattice 18X18 - thermal Power 3850 MW) more than 500 EOC axial shapes from cycle 5 up: - More than 170 shapes for an initial fuel enrichment of 3.5 wt.-% 235-U, discharge burnup ranges from 16.3 to 44.4 MWd/kg; - more than 170 shapes for an initial fuel enrichment of 3.8 wt.-% 235-U, discharge burnup ranges from 14.0 to 52.8 MWd/kg; - more than 180 shapes for an initial fuel enrichment of 4.0 wt.-% 235-U. discharge burnup ranges from 15.5 to 48.9 MWd/kg. PWR AXBUPRO

Two-stage revision of septic knee prosthesis with articulating knee spacers yields better infection eradication rate than one-stage or two-stage revision with static spacers.

Science.gov (United States)

Romanò, C L; Gala, L; Logoluso, N; Romanò, D; Drago, L

2012-12-01

The best method for treating chronic periprosthetic knee infection remains controversial. Randomized, comparative studies on treatment modalities are lacking. This systematic review of the literature compares the infection eradication rate after two-stage versus one-stage revision and static versus articulating spacers in two-stage procedures. We reviewed full-text papers and those with an abstract in English published from 1966 through 2011 that reported the success rate of infection eradication after one-stage or two-stage revision with two different types of spacers. In all, 6 original articles reporting the results after one-stage knee exchange arthoplasty (n = 204) and 38 papers reporting on two-stage revision (n = 1,421) were reviewed. The average success rate in the eradication of infection was 89.8% after a two-stage revision and 81.9% after a one-stage procedure at a mean follow-up of 44.7 and 40.7 months, respectively. The average infection eradication rate after a two-stage procedure was slightly, although significantly, higher when an articulating spacer rather than a static spacer was used (91.2 versus 87%). The methodological limitations of this study and the heterogeneous material in the studies reviewed notwithstanding, this systematic review shows that, on average, a two-stage procedure is associated with a higher rate of eradication of infection than one-stage revision for septic knee prosthesis and that articulating spacers are associated with a lower recurrence of infection than static spacers at a comparable mean duration of follow-up. IV.

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

Modeling the effect of spacers and biofouling on forward osmosis performance

KAUST Repository

Mosqueira Santillá n, Marí a José

2014-01-01

and operational conditions. For this, a two dimensional numerical model for FO systems was developed using computational fluid dynamics (CFD). This model allowed the evaluation of the impact of (i) spacers and (ii) biofilm, and (iii) the combined impact of spacers

Articulating spacers used in two-stage revision of infected hip and knee prostheses abrade with time.

Science.gov (United States)

Fink, Bernd; Rechtenbach, Annett; Büchner, Hubert; Vogt, Sebastian; Hahn, Michael

2011-04-01

Articulating spacers used in two-stage revision surgery of infected prostheses have the potential to abrade and subsequently induce third-body wear of the new prosthesis. We asked whether particulate material abraded from spacers could be detected in the synovial membrane 6 weeks after implantation when the spacers were removed for the second stage of the revision. Sixteen hip spacers (cemented prosthesis stem articulating with a cement cup) and four knee spacers (customized mobile cement spacers) were explanted 6 weeks after implantation and the synovial membranes were removed at the same time. The membranes were examined by xray fluorescence spectroscopy, xray diffraction for the presence of abraded particles originating from the spacer material, and analyzed in a semiquantitative manner by inductively coupled plasma mass spectrometry. Histologic analyses also were performed. We found zirconium dioxide in substantial amounts in all samples, and in the specimens of the hip synovial lining, we detected particles that originated from the metal heads of the spacers. Histologically, zirconium oxide particles were seen in the synovial membrane of every spacer and bone cement particles in one knee and two hip spacers. The observations suggest cement spacers do abrade within 6 weeks. Given the presence of abrasion debris, we recommend total synovectomy and extensive lavage during the second-stage reimplantation surgery to minimize the number of abraded particles and any retained bacteria.

Improvements in nuclear fuel assembly sleeves

International Nuclear Information System (INIS)

Eaton, C.W.; Seeley, T.A.; Ince, G.; Speakman, W.T.

1986-01-01

The graphite sleeve of a nuclear fuel assembly or reflector element for a stringer mounts a number of grids via mounting assemblies installed in grooves formed in the interior wall surface of the sleeve. The bore of the sleeve is of reduced cross-section between two successive grooves such that the internal diameter of the sleeve is substantially the same as the inner diameter of the radially innermost extremity of the mounting assemblies whereby the coolant pressure loss at each transition between the reduced diameter bore section and the mounting assemblies is reduced. Each mounting assembly may be of radially contractable split ring construction to permit its placement in the groove and may carry burnable poison material. (author)

Improvements in nuclear fuel assembly sleeves

Energy Technology Data Exchange (ETDEWEB)

Eaton, C.W.; Seeley, T.A.; Ince, G.; Speakman, W.T.

1986-02-26

The graphite sleeve of a nuclear fuel assembly or reflector element for a stringer mounts a number of grids via mounting assemblies installed in grooves formed in the interior wall surface of the sleeve. The bore of the sleeve is of reduced cross-section between two successive grooves such that the internal diameter of the sleeve is substantially the same as the inner diameter of the radially innermost extremity of the mounting assemblies whereby the coolant pressure loss at each transition between the reduced diameter bore section and the mounting assemblies is reduced. Each mounting assembly may be of radially contractable split ring construction to permit its placement in the groove and may carry burnable poison material.

Implementation of spacer therapy for acute asthma in children.

LENUS (Irish Health Repository)

Vandeleur, M

2009-09-01

The aim was to develop and implement an evidence based guideline for the treatment of acute asthma using a metered dose inhaler and spacer combination. Children admitted to Cork University Hospital Paediatric Department with acute asthma were identified during two identical 2 month seasonal periods before (2005) and after (2006) implementation of the new guidelines in September 2006. Pre-intervention and post-intervention audits by case note review were performed to determine the impact of and compliance with this evidence-based guideline emphasising patient assessment, spacer delivered bronchodilator and specific discharge criteria. Patients had similar characteristics during the two study periods. There was a raised threshold for admission after guideline implementation with 11\\/52 patients having mild exacerbations in 2006, compared to 21\\/36 in 2005. Duration of admission was less in the post-implementation group for equivalent exacerbation severity e.g. for moderate severity; 28 hours in 2005, 23 hours in 2006. Duration of bronchodilator therapy was shorter in 2006 and more likely to be given by spacer device earlier for equivalent levels of severity e.g. for moderate exacerbations, in 2006 the average length of salbutamol therapy was 18 hours with 12 hours by spacer device, in 2005 the average length of therapy was 25 hours with 3 hours by spacer. There was earlier initiation of oral corticosteroids; the average time to administration was 56 minutes in 2006 and 227 minutes in 2005. There was an improved documentation of asthma education in 2006 e.g. inhaler technique was reviewed in 37\\/52 in 2006, 21\\/35 in 2005 and better use of written action plans.

Effect of a spacer moiety on radiometal labelled Neurotensin derivatives

Energy Technology Data Exchange (ETDEWEB)

Mascarin, A.; Valverde, I.E.; Mindt, T.L. [Univ. of Basel Hospital (Switzerland). Div. of Radiopharmaceutical Chemistry

2013-07-01

The binding sequence of the regulatory peptide Neurotensin, NT(8-13), represents a promising tumour-specific vector for the development of radiopeptides useful in nuclear oncology for the diagnosis (imaging) and therapy of cancer. A number of radiometal-labelled NT(8-13) derivatives have been reported, however, the effect of the spacer which connects the vector with the radiometal complex has yet not been investigated systematically. Because a spacer moiety can influence potentially important biological characteristics of radiopeptides, we synthesized three [DOTA({sup 177}Lu)]-X-NT(8-13) derivatives and evaluated the effect of a spacer (X) on the physico-chemical properties of the conjugate including lipophilicity, stability, and in vitro receptor affinity and cell internalization. (orig.)

Bottom nozzle of a LWR fuel assembly

International Nuclear Information System (INIS)

Leroux, J.C.

1991-01-01

The bottom nozzle consists of a transverse element in form of box having a bending resistant grid structure which has an outer peripheral frame of cross-section corresponding to that of the fuel assembly and which has walls defining large cells. The transverse element has a retainer plate with a regular array of openings. The retainer plate is fixed above and parallel to the grid structure with a spacing in order to form, between the grid structure and the retainer plate a free space for tranquil flow of cooling water and for debris collection [fr

Evaluation of the magnitude and effects of bundle duct interaction in fuel assemblies at developmental plant conditions

International Nuclear Information System (INIS)

Serell, D.C.; Kaplan, S.

1980-09-01

Purpose of this evaluation is to estimate the magnitude and effects of irradiation and creep induced fuel bundle deformations in the developmental plant. This report focuses on the trends of the results and the ability of present models to evaluate the assembly temperatures in the presence of bundle deformation. Although this analysis focuses on the developmental plant, the conclusions are applicable to LMFBR fuel assemblies in general if they have wire spacers

Subchannel analysis and correlation of the Rod Bundle Heat Transfer (RBHT) steam cooling experimental data

Energy Technology Data Exchange (ETDEWEB)

Riley, M.P.; Mohanta, L.; Miller, D.J.; Cheung, F.B. [Pennsylvania State Univ., University Park, PA (United States); Bajorek, S.M.; Tien, K.; Hoxie, C.L. [U.S. Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

2016-07-15

A subchannel analysis of the steam cooling data obtained in the Rod Bundle Heat Transfer (RBHT) test facility has been performed in this study to capture the effect of spacer grids on heat transfer. The RBHT test facility has a 7 x 7 rod bundle with heater rods and with seven spacer grids equally spaced along the length of the rods. A method based on the concept of momentum and heat transport analogy has been developed for calculating the subchannel bulk mean temperature from the measured steam temperatures. Over the range of inlet Reynolds number, the local Nusselt number was found to exhibit a minimum value between the upstream and downstream spacer grids. The presence of a spacer grid not only affects the local Nusselt number downstream of the grid but also affects the local Nusselt number upstream of the next grid. A new correlation capturing the effect of Reynolds number on the local flow restructuring downstream as well as upstream of the spacer grids was proposed for the minimum Nusselt number. In addition, a new enhancement factor accounting for the effects of the upstream as well as downstream spacer grids was developed from the RBHT data. The new enhancement factor was found to compare well with the data from the ACHILLLES test facility.

On the development of a grid-enhanced single-phase convective heat transfer correlation

International Nuclear Information System (INIS)

Miller, D.J.; Cheung, F.B.; Bajorek, S.M.

2011-01-01

A new single-phase convective heat transfer augmentation correlation has been developed using single phase steam cooling experimental data obtained from the Penn State/NRC Rod Bundle Heat Transfer (RBHT) facility. Experimental data obtained from the RBHT single phase steam cooling tests have been evaluated and new findings identified. Previous rod bundle tests showed the importance of spacer grid on the local heat transfer, and that the augmentation in heat transfer downstream of a grid decays exponentially. The RBHT data also shows that the Reynolds number affects the rate at which this augmentation decays. The new correlation includes the strong dependence of heat transfer on both the Reynolds number and the grid blockage ratio. While the effects of both parameters were clearly evident in the RBHT experimental data, existing correlations do not account for the Reynolds number effect. The developed correlation incorporates Reynolds number in the decay curve of heat transfer. The newly developed correlation adequately accounts for the dependence of the heat transfer augmentation decay rate on the local flow Reynolds number. (author)

Peripheral pin alignment system for fuel assemblies

International Nuclear Information System (INIS)

Anthony, A.J.

1981-01-01

An alignment system is provided for nuclear fuel assemblies in a nuclear core. The core support structure of the nuclear reactor includes upwardly pointing alignment pins arranged in a square grid and engage peripheral depressions formed in the lateral periphery of the lower ends of each of the fuel assemblies of the core. In a preferred embodiment, the depressions are located at the corners of the fuel assemblies so that each depression includes one-quarter of a cylindrical void. Accordingly, each fuel assembly is positioned and aligned by one-quarter of four separate alignment pins which engage the fuel assemblies at their lower exterior corners. (author)

THYC qualification on Vatican-1 low pressure tests

International Nuclear Information System (INIS)

Duval, C.; Guichard, J.

1991-06-01

PWR cores or fuel assemblies are components of a nuclear power plant involving single and two-phase flows in rod bundles. The knowledge of the detailed two-phase and three-dimensional flow patterns is necessary to evaluate the singularity (grids) and bypass effects on the Departure from Nucleate Boiling (DNB) in reactor cores during incidental transients. For that purpose, since 1989, the VATICAN experiment has been performed at EDF as a part of the qualification program of the three-dimensional computer code THYC, developed by EDF. The qualification strategy of the THYC software for PWR cores is the following: assuming the theoretical or experimental knowledge of regular and singular pressure drops and grid turbulence sources in single-phase, pressure drop multipliers and relative velocity in two-phase flow, the VATICAN experiment allows to evaluate the diffusion phenomena in two-phase flow. It provides thermalhydraulic measurements on a mock-up of a part of 900 MWe PWR fuel assembly in single and two-phase flows, with power and quality gradients. The first configuration of the mock-up, with simple spacer grids, is studied (VATICAN-1). The specific effects of mixing spacer grids will be compared to these data through a second configuration. The last void fraction measurements, using a γ-ray technique, performed on VATICAN-1 low pressure tests allowed to qualify a set of closure relations, particularly a model of little two-phase diffusion, adapted to two-phase flows at low pressure (5.0MPa). The qualification of subcooled boiling and diffusion models will continue on next VATICAN and other experimental campaigns [fr

Manipulation of recombination zone by utilizing the donor of electroplex as a spacer

Energy Technology Data Exchange (ETDEWEB)

Lü, Zhaoyue, E-mail: lvzhaoyue@ecust.edu.cn [Department of Physics, School of Science, East China University of Science & Technology, Shanghai 200237 (China); Yin, Yuehong [Department of Applied Physics, Institute of Physical Chemistry, Henan Polytechnic University, Jiaozuo, Henan Province 454000 (China); Xiao, Jing [Physics and Electronic Engineering College, Taishan University, Shandong Province 271021,China (China)

2016-11-15

The emission of electroplex is varied with applied voltages, leading to poor color stability of devices. In this work, the recombination region was investigated in organic light-emitting diodes (OLEDs) based on the electroplex between N,N’-diphenyl-N,N’-bis(1-naphthyl-phenyl)-1,1′-biphenyl-4,4′-diamine(NPB) and 2-(4-biphenylyl)-5-(4-tert-butylphenyl)-1,2,4-oxadiazole (PBD). Color stability was manipulated by introducing the donor of electroplex as a spacer, which plays an important role of recombination zone for electroplex-based OLEDs. The 5-nm-spacer-device brought out a tolerable chromaticity coordinate shift of less than 0.01 for both x- and y-value. White light emission with CIE coordinates of (0.37,0.28) and (0.33,0.26) at 12 V was achieved when the thickness of spacer was 7 and 10 nm, respectively. The 7-nm-spacer-device exhibited better color stability than the 10-nm-spacer-device. This proves that adjusting spacer thickness is a simple and efficient method to control recombination region and can be useful for fulfilling color-stable electroplex-based white emission.

Manipulation of recombination zone by utilizing the donor of electroplex as a spacer

International Nuclear Information System (INIS)

Lü, Zhaoyue; Yin, Yuehong; Xiao, Jing

2016-01-01

The emission of electroplex is varied with applied voltages, leading to poor color stability of devices. In this work, the recombination region was investigated in organic light-emitting diodes (OLEDs) based on the electroplex between N,N’-diphenyl-N,N’-bis(1-naphthyl-phenyl)-1,1′-biphenyl-4,4′-diamine(NPB) and 2-(4-biphenylyl)-5-(4-tert-butylphenyl)-1,2,4-oxadiazole (PBD). Color stability was manipulated by introducing the donor of electroplex as a spacer, which plays an important role of recombination zone for electroplex-based OLEDs. The 5-nm-spacer-device brought out a tolerable chromaticity coordinate shift of less than 0.01 for both x- and y-value. White light emission with CIE coordinates of (0.37,0.28) and (0.33,0.26) at 12 V was achieved when the thickness of spacer was 7 and 10 nm, respectively. The 7-nm-spacer-device exhibited better color stability than the 10-nm-spacer-device. This proves that adjusting spacer thickness is a simple and efficient method to control recombination region and can be useful for fulfilling color-stable electroplex-based white emission.

How to use an inhaler - no spacer

Science.gov (United States)

... MDI) administration - no spacer; Bronchial nebulizer; Wheezing - nebulizer; Reactive airway - nebulizer; COPD - ... 66. National Asthma Education and Prevention Program website. How to use a metered-dose inhaler. ...

Characterization of feed channel spacer performance using geometries obtained by X-ray computed tomography

KAUST Repository

Haaksman, Viktor A.

2016-09-09

Spiral-wound membrane modules used in water treatment for water reuse and desalination make use of spacer meshes for keeping the membrane leaves apart and for enhancing the mass transfer. Computational fluid dynamics (CFD) has gained importance in the design of new spacers with optimized hydrodynamic characteristics, but this requires a precise description of the spacer geometry. This study developed a method to obtain accurate three-dimensional (3-D) geometry representations for any given spacer design from X-ray computed tomography (CT) scans. The method revealed that the filaments of industrial spacers have a highly variable cross-section size and shape, which impact the flow characteristics in the feed channel. The pressure drop and friction factors were calculated from numerical simulations on five commercially available feed spacers used in practice. Model solutions compared well to experimental data measured using a flow cell for average velocities up to 0.2 m/s, as used in industrial reverse osmosis and nanofiltration membrane operations. A newly-proposed spacer geometry with alternating strand thickness was tested, which was found to yield a lower pressure drop while being highly efficient in converting the pumping power into membrane shear. Numerical model solutions using CFD with geometries from CT scans were closer to measurements than those obtained using the traditional circular cross-section strand simplification, indicating that CT scans are very well suitable to approximate real feed spacer geometries. By providing detailed insight on the spacer filament shape, CT scans allow better quantification of local distribution of velocity and shear, possibly leading to more accurate estimations of fouling and concentration polarization. © 2016 Elsevier B.V.

Effect of plastic spacer handling on salbutamol lung deposition in asthmatic children

DEFF Research Database (Denmark)

Lipworth, Brian J; Lee, Daniel K C; Anhøj, Jacob

2002-01-01

AIMS: To study the effects of electrostatics in a plastic spacer on the lung deposition of salbutamol in asthmatic children. METHODS: Twenty-five children (5-12 years) with mild asthma were given salbutamol hydrofluoroalkane pressurized metered dose inhaler 400 micro g via a 750 ml plastic spacer...... on separate days. Blood samples were taken for plasma salbutamol at 5, 10, 15 and 20 min after inhalation to measure lung bioavailability as a surrogate for relative lung dose. With immediate inhalation following actuation, a new rinsed spacer (NewRinsed ) was compared with a used spacer after repeated daily...... log transformed and expressed as geometric mean fold difference for the average plasma salbutamol concentration (Cav) over 20 min. RESULTS: There were significant differences (P NewRinsed 1.92 fold (95% CI 1.15, 3...

Shelf life of pie caps with biodegradable films as spacers

Directory of Open Access Journals (Sweden)

Daniela Verónica Escobar Gianni

2013-01-01

Full Text Available Commonly pie caps at market use polyethylene films as spacers between them. This paper studies the conventional spacers replacement with edible and biodegradable films made with whey protein isolate (WPI and potassium sorbate as a preservative. Besides facilitating the separation of pie caps, with this application is intended to increase their shelf life. The films made by the compression molding method were used as spacers in pie caps without preservative in their formula (A and with preservative (B and they were compared with conventional polyethylene spacers (C. During four months, monthly sensory, microbiological and physicochemical (humidity evaluations were done on the pie caps, together with humidity and solubility evaluations of the films. None of the samples showed microbiological or sensory deterioration. The sensory attributes showed no or slight difference in study time. Between samples the differences were minor: the best scores were for sample A in color, sample C in flavor, and samples B and C in texture and overall liking. The edible films have an interesting potential for this application, although studies in disguise the flavor of serum should be done.

Hot fuel examination facility element spacer wire-wrap machine

International Nuclear Information System (INIS)

Tobias, D.A.; Sherman, E.K.

1989-01-01

Nondestructive examinations of irradiated experimental fuel elements conducted in the Argonne National Laboratory Hot Fuel Examination Facility/North (HFEF/N) at the Idaho National Engineering Laboratory include laser and contact profilometry (element diameter measurements), electrical eddy-current testing for cladding and thermal bond defects, bow and length measurements, neutron radiography, gamma scanning, remote visual exam, and photography. Profilometry was previously restricted to spiral profilometry of the element to prevent interference with the element spacer wire wrapped in a helix about the Experimental Breeder Reactor II (EBR-II)-type fuel element from end to end. By removing the spacer wire prior to conducting profilometry examination, axial profilometry techniques may be used, which are considerably faster than spiral techniques and often result in data acquisition more important to experiment sponsors. Because the element must often be reinserted into the nuclear reactor (EBR-II) for additional irradiation, however, the spacer wire must be reinstalled on the highly irradiated fuel element by remote means after profilometry of the wireless elements. The element spacer wire-wrap machine developed at HFEF is capable of helically wrapping fuel elements with diameters up to 1.68 cm (0.660 in.) and 2.44-m (96-in.) lengths. The machine can accommodate almost any desired wire pitch length by simply inserting a new wrapper gear module

Knitting Technologies And Tensile Properties Of A Novel Curved Flat-Knitted Three-Dimensional Spacer Fabrics

Directory of Open Access Journals (Sweden)

Li Xiaoying

2015-09-01

Full Text Available This paper introduces a knitting technique for making innovative curved three-dimensional (3D spacer fabrics by the computer flat-knitting machine. During manufacturing, a number of reinforcement yarns made of aramid fibres are inserted into 3D spacer fabrics along the weft direction to enhance the fabric tensile properties. Curved, flat-knitted 3D spacer fabrics with different angles (in the warp direction were also developed. Tensile tests were carried out in the weft and warp directions for the two spacer fabrics (with and without reinforcement yarns, and their stress–strain curves were compared. The results showed that the reinforcement yarns can reduce the fabric deformation and improve tensile stress and dimensional stability of 3D spacer fabrics. This research can help the further study of 3D spacer fabric when applied to composites.

Pros and cons of symmetrical dual-k spacer technology in hybrid FinFETs

Science.gov (United States)

Pradhan, K. P.; Andrade, M. G. C.; Sahu, P. K.

2016-12-01

The symmetrical dual-k spacer technology in hybrid FinFETs has been widely explored for better electrostatic control of the fin-based devices in nanoscale region. Since, high-k tangible spacer materials are broadly became a matter of study due to their better immunity to the short channel effects (SCEs) in nano devices. However, the only cause that restricts the circuit designers from using high-k spacer is the unreasonable increasing fringing capacitances. This work quantitatively analyzed the benefits and drawbacks of considering two different dielectric spacer materials symmetrically in either sides of the channel for the hybrid device. From the demonstrated results, the inclusion of high-k spacer predicts an effective reduction in off-state leakage along with an improvement in drive current. However, these devices have paid the cost in terms of a high total gate-to-gate capacitance (Cgg) that consequently results poor cutoff frequency (fT) and delay.

Using cathode spacers to minimize reactor size in air cathode microbial fuel cells

KAUST Repository

Yang, Qiao

2012-04-01

Scaling up microbial fuel cells (MFCs) will require more compact reactor designs. Spacers can be used to minimize the reactor size without adversely affecting performance. A single 1.5mm expanded plastic spacer (S1.5) produced a maximum power density (973±26mWm -2) that was similar to that of an MFC with the cathode exposed directly to air (no spacer). However, a very thin spacer (1.3mm) reduced power by 33%. Completely covering the air cathode with a solid plate did not eliminate power generation, indicating oxygen leakage into the reactor. The S1.5 spacer slightly increased columbic efficiencies (from 20% to 24%) as a result of reduced oxygen transfer into the system. Based on operating conditions (1000ς, CE=20%), it was estimated that 0.9Lh -1 of air would be needed for 1m 2 of cathode area suggesting active air flow may be needed for larger scale MFCs. © 2012 Elsevier Ltd.

Experimental determination of temperature fields in sodium-cooled rod bundles with hexagonal rod arrangement and grid spacers

International Nuclear Information System (INIS)

Moeller, R.; Tschoeke, H.; Kolodziej, M.

1977-01-01

Three-dimensional temperature fields in the claddings of sodium cooled rods were determined experimentally under representative nominal operating conditions for a SNR typical 19-rod bundle model provided with spark-eroded spacers. These experiments are required to verify thermohydraulic computer programs which will provide the output data for strength calculations of the high loaded cladding tubes. In this work the essentials are reported of the measured circumferential distributions of wall temperatures of peripheral rods. In addition the sub-channel temperatures measured over the bundle cross section are indicated, they are required to sustain codes for the global thermohydraulic design of core elements. The most important results are: 1) The whole fuel element is located within the thermal entrance length. 2) High azimuthal temperature differences were measured in the claddings of peripheral rods, which are strongly influenced by the distance between the rod and the shroud, especially for the corner rod. 3) With decreasing Pe-number ( [de

Nuclear fuel assembly incorporating primary and secondary structural support members

International Nuclear Information System (INIS)

Carlson, W.R.; Gjertsen, R.K.; Miller, J.V.

1987-01-01

A nuclear fuel assembly, comprising: (a) an upper end structure; (b) a lower end structure; (c) elongated primary structural members extending longitudinally between and rigidly interconnecting the upper and lower end structures, the upper and lower end structures and primary structural members together forming a rigid structural skeleton of the fuel assembly; (d) transverse grids supported on the primary structural members at axially spaced locations therealong between the upper and lower end structures; (e) fuel rods extending through and supported by the grids between the upper and lower end structures so as to extend in generally side-by-side spaced relation to one another and to the primary structural members; and (f) elongated secondary structural members extending longitudinally between but unconnected with the upper and lower end structures, the secondary structural members extending through and rigidly interconnected with the grids to extend in generally side-by-side spaced relation to one another, to the fuel rods and to the primary structural members so as to bolster the stiffness of the structural skeleton of the fuel assembly