Modeling In-Reactor Deformation of Zr-2.5Nb Pressure Tubes in CANDU Power Reactors PDF Download
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Author: N. Badie Publisher: ISBN: Category : Congress Languages : en Pages : 20
Book Description
Changes in shape of internally pressurized tubes caused by operating temperatures and pressures are enhanced by fast neutron irradiation. Lengths and diameters of Zr-2.5Nb pressure tubes in CANada Deuterium Uranium-Pressurized Heavy Water (CANDU-PHW) power reactors and test reactors have been monitored periodically over the past 20 years. Axial and transverse strain rates have been evaluated in terms of operating variables and the crystallographic texture and anisotropic microstructure of the extruded and cold-drawn tubes. The anisotropic deformation occurring during steady-state irradiation creep and growth is described by a self-consistent model that takes into account the presence of intergranular stresses without building up any discontinuities of strain and stress at the grain boundaries. In this model, it is assumed that climb-assisted glide of dislocations on prismatic, basal, and pyramidal planes is the dominant creep mode and that growth occurs by net fluxes of interstitials and vacancies to a non-random distribution of dislocations and grain boundaries. The predictions from a deformation equation based on data from the Pickering and Point Lepreau Nuclear Generating Stations and the WR1, Osiris, DIDO, and NRU test reactors are in good agreement with measurements of pressure tubes in Bruce units. The equation has been employed as a material subroutine in the 3-D finite element code H3DMAP for predicting the detailed shape change of pressure tubes. The prediction from H3DMAP is a more complete description of shape change than that obtained from the closed-form expression.
Author: N. Badie Publisher: ISBN: Category : Congress Languages : en Pages : 20
Book Description
Changes in shape of internally pressurized tubes caused by operating temperatures and pressures are enhanced by fast neutron irradiation. Lengths and diameters of Zr-2.5Nb pressure tubes in CANada Deuterium Uranium-Pressurized Heavy Water (CANDU-PHW) power reactors and test reactors have been monitored periodically over the past 20 years. Axial and transverse strain rates have been evaluated in terms of operating variables and the crystallographic texture and anisotropic microstructure of the extruded and cold-drawn tubes. The anisotropic deformation occurring during steady-state irradiation creep and growth is described by a self-consistent model that takes into account the presence of intergranular stresses without building up any discontinuities of strain and stress at the grain boundaries. In this model, it is assumed that climb-assisted glide of dislocations on prismatic, basal, and pyramidal planes is the dominant creep mode and that growth occurs by net fluxes of interstitials and vacancies to a non-random distribution of dislocations and grain boundaries. The predictions from a deformation equation based on data from the Pickering and Point Lepreau Nuclear Generating Stations and the WR1, Osiris, DIDO, and NRU test reactors are in good agreement with measurements of pressure tubes in Bruce units. The equation has been employed as a material subroutine in the 3-D finite element code H3DMAP for predicting the detailed shape change of pressure tubes. The prediction from H3DMAP is a more complete description of shape change than that obtained from the closed-form expression.
Author: N. Christodoulou Publisher: ISBN: Category : Languages : en Pages : 0
Book Description
Changes in shape of internally pressurized tubes caused by operating temperatures and pressures are enhanced by fast neutron irradiation. lengths and diameters of zr-2.5nb pressure tubes in canada deuterium uranium-pressurized heavy water (candu-phw) power reactors and test reactors have been monitored periodically over the past 20 years. axial and transverse strain rates have been evaluated in terms of operating variables and the crystallographic texture and anisotropic microstructure of the extruded and cold-drawn tubes. the anisotropic deformation occurring during irradiation creep and growth is described by a self-consistent model that takes into account the presence of intergranular stresses without building up any discontinuities of strain and stress at the grain boundaries. in this model it is assumed that climb assisted glide of dislocations on prismatic, basal and pyramidal planes is the dominant creep mode and that growth occurs by net fluxes of interstitials and vacancies to a non random distribution of dislocations and grain boundaries. the predictions from a deformation equation based on data from the pickering, and point lepreau nuclear generating stations and the wr1, osiris, dido and nru test reactors are in good agreement with measurements of pressure tubes in bruce units. the equation has been employed as a material subroutine in the 3-d finite element code h3dmap for predicting the detailed shape change of pressure tubes. the prediction from h3dmap is a more complete description of shape change than that obtained from the closed-form expression.
Author: AR. Causey Publisher: ISBN: Category : Deformation Languages : en Pages : 15
Book Description
Changes in shape of internally pressurized tubes caused by operating temperatures and pressures are enhanced by fast neutron irradiation. Lengths and diameters of Zr-2.5 wt% Nb pressure tubes in CANada Deuterium Uranium-Pressurized Heavy Water (CANDU-PHW) power reactors and test reactors have been monitored periodically over the past 15 years. Axial and transverse strain rates have been evaluated in terms of the operating variables and the crystallographic texture and anisotropic microstructure of the extruded and cold-drawn tubes. The anisotropic deformation can be described by models for creep and irradiation growth in which the anisotropy factors are calculated from texture. It is assumed that prismatic slip is the dominant creep mode and that growth occurs by net fluxes of interstitials to a nonrandom distribution of ?a? type edge dislocations and vacancies to ?a? type screw dislocations, ?c? type edge dislocations, and grain boundaries. The equations based on data from the Pickering Generating Station and WR1 test reactors give good agreement with measurements on internally pressurized tubes in Bruce Generating Station and the National Reactor Universal (NRU) test reactor and uniaxially stressed specimen in NRU.
Author: A. R. Causey Publisher: Chalk River, Ont. : Reactor Materials Research Branch, Chalk River Laboratories ISBN: 9780660151892 Category : Nuclear reactors Languages : en Pages : 22
Book Description
The anistropy of creep deformation of Zr-2.5Nb pressure tubes during service in CANDU reactors is related to the anisotropic physical properties of the hexagonal crystal structure of zirconium. These physical properties contribute to the development during fabrication of an anisotropic microstructure, including crystallographic textures, grain morphologies, and dislocation structures. A number of studies tried to relate the anisotropic deformation of the polycrystalline zirconium alloys to those of their individual grains by accounting for the microstructural features, particularly the crystallographic texture, but they suffered from a lack of experimental data from biaxial creep tests on materials that have crystallographic texture similar to that of the pressure tubes. This experiment contributes to the development of a reliable model for Zr-2.5Nb tubes by using two batches of small tubes, one of which has a crystallographic texture similar to that of the CANDU power reactor pressure tubing, the other having a texture that is completely different. The results are analyzed in terms of texture using a self-consistent model to account for the effects of the grain interactions.
Author: M. Griffiths Publisher: ISBN: Category : Climb Languages : en Pages : 9
Book Description
The diametral expansion, elongation, and sag rates of Zr-2.5Nb pressure tubes in CANDU® (CANada Deuterium Uranium) nuclear reactors are important properties that limit their useful life and the maximum power level for reactor operation. As a result irradiation creep models are needed to predict the deformation behavior of the core components over the reactor life. It is important to know the creep behavior as a function of neutron flux in order to develop creep models over the range of operating conditions in the reactor core. At the edge of the reactor core, the neutron flux is decreasing very rapidly and there is a complex transition in creep behavior from irradiation-dominated creep to thermal-dominated creep. Also, mechanical properties such as tensile strength, fracture toughness, and delayed hydride-cracking are changing in the transition from thermal to irradiation conditions at the edge of the reactor core. Detailed studies have been completed on a Zr-2.5Nb tube irradiated in the NRU materials test reactor at Chalk River Laboratories. Pressure tube 601 was operating for a period of 66 950 h at temperatures ranging from about 547 K at the inlet and 571 K at the outlet. After the tube was removed in 1988 samples were taken for retrospective dosimetry to determine the fast neutron flux along the assembly. It was determined that the tube had been irradiated to a peak fluence of about 6x1025 n.m-2 corresponding to a fast neutron flux of about 2x1017 n.m-2.s-1. The flux profile was mapped and it was clear that the flux dropped rapidly to negligible values at about 0.5 m from the ends of the fueled zone. Samples of pressure tubes were taken for hardness testing and characterization by TEM and XRD analysis at various locations corresponding with different operating conditions (neutron flux and temperature) but at the same time. The creep behavior during operation was obtained by periodic gaging of the pressure tube internal diameter. The results of the microstructure characterization are presented and discussed in relation to the measured mechanical properties (creep and hardness). The microstructure and mechanical properties change significantly in the transition from the unirradiated state up to fluxes of about 1x1017 n.m-2.s-1.
Author: G. A. Bickel Publisher: ISBN: Category : Crystallographic texture Languages : en Pages : 22
Book Description
In an Advanced CANDU Reactor (ACR) (ACR is a registered trademark of Atomic Energy of Canada Limited), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure, and flux than those in a CANDU (CANDU is a registered trademark of Atomic Energy of Canada Limited) reactor, and accordingly require a thicker wall (6.5 mm for ACR versus 4.2 mm for CANDU). In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a study to model and forecast the performance of these thicker pressure tubes has been undertaken. One of the main requirements for the pressure tube is to have low diametral creep. Based on previous experience with CANDU reactor pressure tube performance and manufacture, an assessment of the grain structure and texture of the ACR pressure tubes indicates that the in-reactor creep deformation will be improved. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor.
Author: H. Z. Fan Publisher: ISBN: Category : Languages : en Pages : 0
Book Description
The objective of this study is to evaluate the relative impact of the design parameters on bundle uranium mass and sheath strain and to re-evaluate the basis for the limitation on bundle mass due to an increase in bundle subchannel cross-sections. bundle uranium mass is determined by parameters that in turn affect the sheath strain during operation. this might affect sub-channel flow areas and affect the chf-ccp. the bundle uranium mass was assessed with electres and resulting sheath strains estimated for a candu 6 fuel channel operating at overpowers just at the trip set point of the reactor (onset of sheath dryout), a 14% power increase. the electres fuel modeling code is used to determine the relative impact on sheath strain of the design parameters that control uranium mass, namely, pellet density, diametral clearance, axial gap, and pellet face geometry (chamfer, dish depth, and land width). a limitation was placed on bundle uranium mass by new brunswick power. this came from a ccp evaluation showing that a candu 6 reactor, fuelled with bundles having average masses greater than 19.25 kg u, would have a net positive sheath strain over a fuel channel at the power for the onset of dryout, and therefore a ccp penalty. the calculations were based on steady bundle powers, operating in a fuel channel at ccp to a burnup of 168 mw middle dot h/kg u. at this burnup the strain calculation included a 14% power boost. these are indeed very conservative assumptions with a view to maximizing calculated sheath strains, without regard for fuel defect probability. for comparison, this study has produced electres strain calculations for high power channel power histories representative of 8 bundle shifts, also with a 14% power boost, operating at dryout.
Author: N. Badie
Author: N. Badie
Author: N. Christodoulou
Author: N. Christodoulou
Author: AR. Causey
Author: M. Griffiths
Author: A. R. Causey
Author: M. Griffiths
Author: G. A. Bickel
Author: N. Christodoulou
Author: H. Z. Fan