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Irradiation creep strain due to climb-controlled glide process calculated by means of point defect kinetics theory at 823 K and 100 MPa: (a) in the case that void number density increases linearly with neutron dose, and (b) in the case that void number density is constant
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The irradiation creep-swelling interaction parameters were pecisely derived for MONJU fuel pin cladding PNC316 by irradiation tests of pressurized tubes in FFTF. It was found out that a creep-swelling coupling coefficient decreased and asymptotically approached a constant value as the swelling progresses, although it was widely believed that the ir...
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Although the pearlitic steel is one of the most extensively studied materials, there are still questions unanswered about the interface in the lamellar structure. In particular, to deepen the understanding of the mechanical behavior of pearlitic steel with fine lamellar structure, it is essential to reveal the structure-property relationship of the...
Investigation of material creep behaviour in the diffusion controlled creep regime is often unfeasible because of the long duration associated with low stress levels. On the other side, extrapolation from higher creep rates usually provides inaccurate results because of the sharp change in the data trend as a result of the change in the governing d...
Citations
... The swelling rate is a function of the evolving cavity density and the dislocation density, as one might expect. Figure 30 is a reasonable facsimile of real behaviour [95]. It is likely that D represents the strain arising directly from anisotropic diffusional mass transport subject to an applied stress, but it could also include an element of creep that is based on dislocation slip of network dislocations, i.e., stress-induced climb and glide (SICG) that is enhanced by swelling. ...
Austenitic stainless steels are used for core internal structures in sodium-cooled fast reactors (SFRs) and light-water reactors (LWRs) because of their high strength and retained toughness after irradiation (up to 80 dpa in LWRs), unlike ferritic steels that are embrittled at low doses (<1 dpa). For fast reactors, operating temperatures vary from 400 to 550 °C for the internal structures and up to 650 °C for the fuel cladding. The internal structures of the LWRs operate at temperatures between approximately 270 and 320 °C although some parts can be hotter (more than 400 °C) because of localised nuclear heating. The ongoing operability relies on being able to understand and predict how the mechanical properties and dimensional stability change over extended periods of operation. Test reactor irradiations and power reactor operating experience over more than 50 years has resulted in the accumulation of a large amount of data from which one can assess the effects of irradiation on the properties of austenitic stainless steels. The effect of irradiation on the intrinsic mechanical properties (strength, ductility, toughness, etc.) and dimensional stability derived from in- and out-reactor (post-irradiation) measurements and tests will be described and discussed. The main observations will be assessed using radiation damage and gas production models. Rate theory models will be used to show how the microstructural changes during irradiation affect mechanical properties and dimensional stability.
... Predictions for cladding stresses due to fission gas depend on the fuel type and burnup, length of the plenum, and the diameter and wall thickness of the cladding. As an example, a standard FFTF mixed oxide (MOX) fuel pin can develop an effective stress of approximately 40 MPa after about 100 dpa of exposure [16]. Stresses from gas pressure and fuel clad mechanical interaction (FCMI) have Diagram showing typical layout of fuel pin bundle assembly and duct in a fast reactor [6]. ...
... 79 and 80 that doubling the hoop stress did not double the strain rate in the tube. 5. The coupling coefficient D tends to fall to zero rather quickly when swellingbefore-creep occurs but falls more slowly in creep-before-swelling scenarios (fuel pins vs. pressurized tubes [200]. [197]. ...
... A consensus explanation of the creep disappearance phenomena has not yet been reached. Various models have been proposed involving voids acting to erase the anisotropy of dislocation Burgers vector [203,204] and the involvement of precipitate sinks to serve a strong sinks that compete with dislocations [200]. ...
Austenitic stainless steels used as fuel cladding or structural components in various reactor types must often withstand an exceptionally strenuous and challenging environment, even in the absence of neutron irradiation. In addition to the environmental challenges associated with surviving prolonged exposure to high temperature and corrosive media, exposure to displacive neutron irradiation pushes steels into a progressive nonequilibrium evolution involving generation of completely new microstructural components unique to the radiation environment and new phase structures driven by diffusional processes that operate only during irradiation. The result of such microstructural and microchemical alteration is often the production of an alloy that cannot be found on an equilibrium diagram. Concurrent with this evolution on the microscopic scale are macroscopic changes in physical and mechanical properties, and most importantly, changes in dimensional stability that are outside the realm of normal engineering experience. Such changes can strongly affect the structural integrity and limit the in-reactor lifetime of structural steels. While reviews of this type usually focus primarily on the displacive aspects of neutron irradiation and perhaps confine the review to one or two types of reactor, it is necessary in this review to give equal attention to the transmutation aspects of irradiation for the full range of neutron flux-spectral environments. This review first focuses on identifying the important characteristics of various reactor environments, then addresses the microscopic aspects of neutron-induced alteration, and finally moves to cover the macroscopic consequences of these alterations.
... The coupling coefficient D tends to fall to zero rather quickly when swelling-before-creep occurs but falls more slowly in creep-before-swelling scenarios (fuel pins vs. pressurized tubes). 175 A consensus explanation of the creep disappearance phenomena has not yet been reached. Various models have been proposed involving voids acting to erase the anisotropy of dislocation Burgers vector 176,177 and the involvement of precipitate sinks to serve as strong sinks that compete with dislocations. ...
... Various models have been proposed involving voids acting to erase the anisotropy of dislocation Burgers vector 176,177 and the involvement of precipitate sinks to serve as strong sinks that compete with dislocations. 175 As mentioned earlier, once swelling begins, irradiation creep quickly assumes all the parametric dependencies of void swelling. However, for many years it was assumed that the B 0 component of creep was also strongly dependent on dpa rate, increasing as the dpa rate fell, as shown in Figure 81. ...
... One would expect that the swelling-enhanced creep rate would also increase as swelling increases. It therefore appears that attainment of large swelling in each assembly prior to the development of significant gas pressure has almost certainly caused the cladding to reach the ''creep cessation'' or ''creep disappearance'' stage defined by Garner and coworkers, whereby creep initially accelerates with swelling but then ''disappears'' as the swelling rate approaches the 1%/dpa level [16,383940414243. Looking only at Figs. 5–7 it might be reasonable to question the conclusion that the 1%/dpa asymptote justifies a conclusion that all high swelling points on the loop were reached along a 1%/dpa path. ...
Three 37-pin MOX-fueled experimental subassemblies were irradiated in EBR-II with fuel pin cladding constructed from annealed AISI 304 stainless steel. Analysis of the swelling and irradiation creep of the cladding showed that the terminal swelling rate of AISI 304 stainless steel appears to be ∼1%/dpa and that swelling is very reproducible for identical irradiation conditions. The swelling at a given neutron fluence is rather sensitive to both irradiation temperature and especially to the neutron flux, however, with the primary influence residing in the transient regime. As the neutron flux increases the duration of the transient regime is increased in agreement with other recent studies. The duration of the transient regime is also decreased by increasing irradiation temperature.In these assemblies swelling reached high levels rather quickly, reducing the opportunity for fuel pin cladding interaction and thereby reducing the contribution of irradiation creep to the total deformation. It also appears that in this swelling-before-creep scenario that the well-known “creep disappearance” phenomenon was operating strongly.
... burnup, featured less than 2% irradiation creep strain even after 5 times 50% over power transients without rupture. Similar results were provided by Ukai and Ohtsuka (2007). Further, reduction of D9 steel Yield Strength (YS) and Ultimate Tensile Strength (UTS) up to 70 dpa irradiation dose is lower than 15% at 1000 K. Thus, the influence from a higher irradiation dose ($120 dpa) on the creep rupture property of D9 steel may be neglected (Eriksson et al., 2005b, Cannon et al., 1992. ...
We have performed transient analysis of a medium size sodium cooled reactor loaded with different fractions of americium in the fuel. Unprotected Loss of Flow (ULOF) and Unprotected Transient over Power (UTOP) accidents were simulated in a geometrical model of BN600, using safety parameters obtained with the SERPENT Monte Carlo code.For the reference case of MOX fuel, with 1 wt.% Am concentration at BOL and EOEC, the margin to melt following a design basis condition UTOP is about 100 K for a maximum linear rating of 37 kW/m. Introducing Am, the maximum fuel temperature increases by 50 K per percent Am in the fuel. In order to maintain a margin of 100 K to fuel failure, the linear rating has to be reduced by ∼6% for each percent Am added to the fuel.Hence, an Am concentration of 2–3 wt.% in the fuel would lead to a power penalty of 6–13%, permitting a consumption rate of 1.7–3.4 kg Am/TWhth.
... This trend is confirmed by Garner's review and discussion [6] of the collection and interpretation of experimental data from which magnitude of the creep-swelling coupling coefficient D has been derived. In a more recent paper Ukai and Ohtsuka [23] analyzed fuel pin cladding and pressurized tube data obtained on Modified 316 stainless steel and found that D decreased and asymptotically approached a constant value with increasing swelling strain. Using [8]. ...
Irradiation induced swelling of reactor core materials may jeopardize safe and reliable operation of fast reactors due to swelling-induced distortion and interference of core components. The principles of incremental continuum plasticity are used here to develop constitutive equations that can be used to conduct engineering evaluations of these potential problems. The equations are used in Part II to analyze previously unreported in-reactor creep and swelling data obtained ca. 1977–1979 as part of the US breeder reactor program. Results of this stress state experiment showed for the first time that a deviatoric stress can affect volumetric swelling. The constitutive equations developed here predict that, in the presence of significant swelling, deviatoric and volumetric strain rate components each are functions of both deviatoric and hydrostatic components of stress for both linear and non-linear creep.
Garner Frank A. (2020). Radiation-Induced Damage in Austenitic Structural Steels Used in Nuclear Reactors.
In: Konings, Rudy JM and Stoller Roger E (eds.) Comprehensive Nuclear Materials 2nd edition, vol. 3, pp. 57–168. Oxford: Elsevier.
10.1016/B978-0-12-803581-8.12067-3
Irradiation creep was investigated in different austenitic steels. Pressurized tubes with stresses of 127 to 220 MPa were irradiated in BOR-60 at 320°C to 120 dpa. Creep behavior was dependent on both chemical composition and metallurgical state of steels.
Different steels irradiated with and without stress were examined by TEM. Without stress, the irradiation produced high densities of dislocation lines and Frank loops and, depending on the type of steels, precipitates. Stress induced an increase of the precipitate mean size and density and, for some grades, an increase of the mean loop size and a decrease of their density. An anisotropy of Frank loop density or size induced by stress was not observed systematically. Dislocation line microstructure seems not to be different between the stressed and unstressed specimens. No cavities were detectable in these specimens.
By comparing with the data from this work, the main irradiation creep models are discussed.
