Kinetics of Radiation-Induced Segregation in Ternary Alloys

Lam, Nghi Q.; Kumar, Arvind S.; Wiedersich, H.

doi:10.1520/stp34393s

Cited by 19 publications

(16 citation statements)

References 0 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The model WaStested against MS measurements from seven different Fe-Cr-Ni dloYs, ticfiated with protons at 400"C to doses from 0.1 to 3.0 dpa, and at 0.5 dpa at mperaturesfrom ZOO-~OC.~total, over 1100 grain boundary composition measurements were obttined. The~K m~el Wm found to IX superior to other proposed RIS models [7,8], dicting he measured segregation with no error greater than 5 at'%but usually less than 2 at%. 'is 'improvement is due to the fact that the MIK model crdcukitesatom-vacancy migration energies 'don Pair potentials, the ordering energy, and the locaf composition.…”

Section: Introductionmentioning

confidence: 91%

Modeling the Effect of Irradiation and Post-Irradiation Annealing on Grain Boundary Composition in Austenitic Fe-Cr-Ni Alloys

et al. 1998

View full text Add to dashboard Cite

A.wr'RAcr {my irra~ation effects in. Fe-Cr-Nl alloys such as radiation-induced segregation, radiationcnharrceddlffusloq, and void swelhng$are known to vary with btdk alloy composition. The &velopment of~c~stIuctu@ and mtcrochemicaf changes during imadiation and during postitradiationanneal~g ISdetennmed by the rate of diffusion of point defects and alloying elements. Toaccuratelyp@ct the changes m grain boundaty chertristty due to radiation-induced segregation and~st-madlation anneahng. the composition dependence of dif~lon parameters, such as the tnigrmon energy, must be known. A model has been developed which calctdates migration energies using pti~temtion energies~ereby~o~~g for the effect of .comPsition on diffusivity. The advantages of dus CaIculationalmethcd are that a single set of snput parameters cur be used for a wide range of bulk alloy compositions, and the effects of locaf order can easily be incorporatedinto the calculations. A description of the model is presented, and model calculations are compared to segregation measurements from seven different iron-chromium-nickel alloys, imdiated with protons to doses from 0.1 to 3.0 dpa at temperatures between 2CXYC and 600°C. Resultsshow that segregation trends can be modeled using a single set of input parameters with the differencebetween model calculation md measurement being less than 5 at%, but usuafly less than 2 at%. Additionally, model predictions we compared to grain boundary crrm;mi:ion measurementsof neutron irradiated 304 stainless steel foflowing anneahg.For the fimited annealingdata available, model calculations correctly predict the magnitude and time wale for recovery of the grain boundary composition.

show abstract

Section: Introductionmentioning

confidence: 91%

Modeling the Effect of Irradiation and Post-Irradiation Annealing on Grain Boundary Composition in Austenitic Fe-Cr-Ni Alloys

et al. 1998

View full text Add to dashboard Cite

show abstract

“…Models have been developed and calculations performed only near the surface of a sample or grain boundaries (planar sinks for point defects) for concentrated double, triple and dilute binary substitution alloys [19][20][21][22]. Models have been developed and calculations performed only near the surface of a sample or grain boundaries (planar sinks for point defects) for concentrated double, triple and dilute binary substitution alloys [19][20][21][22].…”

Section: Modeling Of Radiation-induced Segregation In Alloysmentioning

confidence: 99%

Change in the Properties of Structural Materials Under Irradiation: Role of Radiation-Induced Segregation

et al. 2014

View full text Add to dashboard Cite

There are several alternative strategies for the development of nuclear power [1,2]. One of the basic requirements imposed on the future nuclear power technology is its large scale, which presupposes a higher level of safety for all its elements ranging from reactor facilities to NFC technologies [3,4]. One direction for organizing a new technological platform for nuclear power, work on which is now being performed at the State Corporation Rosatom, is the development of innovative fast reactors which have extreme temperatures and dose loads and are cooled by sodium, lead and lead-bismuth [5,6]. In this connection, research on the properties of structural materials under irradiation has not lost its urgency.Metals and alloys irradiated by damaging particles become supersaturated with point defects and radiation-accelerated diffusion. As a result of radiation-induced segregation, the composition of the alloys becomes nonuniform near sinks for point defects (the boundaries of the samples and grains, dislocations, loops, pores and segregations). This is caused by two mechanisms: 1) the presence of a concentration gradient of point defects near sinks and the difference of the diffusion mobility of the components of an alloy and 2) the formation of complexes of certain chemical elements of the alloy with point defects and flows of point defects toward sinks.Alloys undergoing prolonged thermal soaking strive toward thermodynamic equilibrium when there are no concentration gradients of the components. Under irradiation, alloys strive toward 'kinetic' equilibrium (a stationary state), where the fluxes of the components equal zero but because of a concentration gradient of point defects concentration gradients of the components of the alloy arise and the alloy exhibits local nonuniformity. The phase composition of the alloy can change considerably as a result of radiation-induced segregation, irrespective of its initial state.Radiation-induced segregation was first predicted theoretically in 1971 [7] and it was first observed in 1973 [8] in the 18Cr-8Ni-1Si alloy irradiated by fast electrons. Subsequently, many experimental and theoretical studies showing its principal regularities in dilute and concentrated model alloys as well as in structural materials have been performed [9][10][11][12][13][14]. It was shown that the temperature range of the phenomenon (0.2-0.6)T melt coincides with that of vacancy swelling of metals and alloys.The data in [13] indicate the following features of radiation-induced segregation in structural materials: austenitic steel -enrichment with Ni, Si, P, and S and depletion of Fe, Cr, Mo, Ti, and Mn in regions near point defects; ferrite-martensite steel -enrichment with Ni, P, and Si in regions near sinks and the data on the behavior of chromium are ambiguous; vessel steel -enrichment with P, Ni, and Si in regions near sinks. It is evident in Fig. 1 that the radiation-induced segregation shifts into regions of higher temperature with higher rate of production of displacements (in dpa/sec).

show abstract

“…RIS RIS describes the redistribution of major alloying elements and impurity elements at point defect sinks (Allen, Busby, Was, & Kenik, 1998;Hindmarsh, 1972;Jenssen et al, 1998;Lam, Kumar, & Wiedersich, 1982;Okamoto & Rehn, 1979;Okamoto & Wiedersich, 1974;Perks, Marwick, & English, 1987; Simonen (personal communication); Was, 2007;Was et al, 1999a;Wiedersich, Okamoto, & Lam, 1979). As IASCC is intergranular, the sinks of greatest interest are the grain boundaries.…”

Section: Metallurgy and Microstructurementioning

confidence: 99%

Irradiation-assisted stress corrosion cracking

Was¹,

Ashida²,

Andresen³

2011

Corrosion Reviews

View full text Add to dashboard Cite

Irradiation-assisted stress corrosion cracking (IASCC) refers to the process(es) by which irradiation enhances the inherent susceptibility of alloys to stress corrosion cracking (SCC). This article begins with an introduction to the IASCC problem experienced in light water reactors, followed by a summary of existing plant and laboratory data that reveal its dependence on irradiation, environment and alloy parameters. The specifi c effects of irradiation on IASCC are classifi ed into two categories: the effects on water chemistry and the effects on microstructure. Water chemistry effects include radiolysis and its effects on corrosion potential, and effects of corrosion potential on IASCC. Microstructure effects include radiationinduced segregation (RIS), irradiated microstructure, swelling and creep, and hydrogen and helium generation. Leading mechanisms proposed to explain the role of radiation in the SCC process are: radiolysis and crack tip strain rate, grain boundary chromium depletion, irradiation hardening, localized deformation and RIS of minor elements. As nuclear power plants operate for longer, an increased incidence of SCC can be expected unless active mitigation steps are taken. Drawing on the accumulated understanding over the past four decades of the key processes believed to control IASCC, a set of attributes of an ideal, IASCC-resistant alloy are identifi ed as a fi rst step.

show abstract

Kinetics of Radiation-Induced Segregation in Ternary Alloys

Cited by 19 publications

References 0 publications

Modeling the Effect of Irradiation and Post-Irradiation Annealing on Grain Boundary Composition in Austenitic Fe-Cr-Ni Alloys

Modeling the Effect of Irradiation and Post-Irradiation Annealing on Grain Boundary Composition in Austenitic Fe-Cr-Ni Alloys

Change in the Properties of Structural Materials Under Irradiation: Role of Radiation-Induced Segregation

Irradiation-assisted stress corrosion cracking

Contact Info

Product

Resources

About