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Features of structure-phase transformations and segregation processes under irradiation of austenitic and ferritic-martensitic steels
EISEVIER
Journal of Nuclear Materials 212-21.5 (1994) 39-44
Features of structure-phase transformations and segregation processes under irradiation of austenitic and ferritic-martensitic steels I.M. Neklyudov, V.N. Voyevodin Kharkov Institute of Physics and Technology, 310108 Kharkou, Ukraine
Abstract The difference between crystal lattices of austenitic and ferritic steels leads to distinctive features in mechanisms of physical-mechanical change. This paper presents the results of investigations of dislocation structure and phase evolution, and segregation phenomena in austenitic and ferritic-martensitic steels and alloys during irradiation with heavy ions in the ESUVI and UT1 accelerators and by neutrons in fast reactors BOR-60 and BN-600. The influence of different factors (including different alloying elements) on processes of structure-phase transformation was studied.
1. Introduction
2. Experimental procedures and materials
Complex interdependent processes take place in solids during their irradiation: nuclear processes resulting in appearance of primary knockon atoms and new elements; atom-atomic cascade collisions, which produce crystal lattice imperfections in the form of point defects, crowdions and depleted zones; substructural processes leading to the formation of directly registered clusters, nuclei of dislocation loops, precipitate particles and voids; diffusional ones which are responsible for evolution of micro- and macrostructures, solute and phase compositions. The substructural and diffusional processes play a major role in physicalmechanical property changes. Stainless austenitic and ferritic-martensitic steels are candidates for structural materials for fusion reactor core applications. The difference in their chemical composition and crystal structure results in distinctive features of the diffusional and substructural processes occurring during irradiation and, consequently, in the radiation resistance of these steels [1,2]. This paper is a brief review of results on some aspects of structure-phase transformations and radiation-induced segregation mechanisms under irradiation of austenitic and ferritic steels.
The chemical composition and thermal treatment of the steels studied are given in Tables 1 and 2. The structure-phase state of initial and irradiated steel specimens was examined with the aid of transmission electron microscopes (TEMs): EM-125K TEM and JEM-IOOCX TEM equipped with ASID-4D adapter and X-ray energy-dispersive spectrometer Link Systems-860. A quantitative element analysis was carried out using the RTS2/FLS program. For irradiations the ESUVI (material research heavy ion accelerator), the UT1 (heavy ion accelerator), and fast reactors BOR-60 and BN-600 were used, with the irradiation conditions given in Table 2.
3. Results and discussion 3.1. Evolution
of dislocation
structure
The radiation resistance of steels is determined mainly by the rate of nucleation and evolution of dislocation loops. The type and parameters of disloca-
0022-3115/94/$07.00 0 1994 Elsevier Science B.V. All rights reserved SSDI 0022-3115(94)00277-U
I.M. Neklyudor~, V.N. Vp~~erndin/Journal
40 Table 1 Chemical composition
18Cr-1ONi 16Cr-15Ni 13Cr-13Ni 16Cr-11Ni 13Cr-2Mo 9Cr-2Mo 26Cr-6Ni
of investigated
materials
(wt%)
Ni
Cr
Mn
MO
C
Ti
Nb
Si
SC
Thermal
10.6 14.5 13.2 10.7 0.2 _ 5.46
18.0 15.5 13.1 16.2 12.1 9.42 25.6
1.2 0.8 2.5 1.3 0.5 0.7 0.8
2.9 0.5 2.3 1.8 1.5 -
0.04 0.04 0.06 0.15 0.13 0.02
0.06 2.7 -
0.8 _ _ 0.25 0.31 -
0.8 0.3 _ 0.6 0.5 0.6 0.6
0.02-0.05 0.05-0.2 0.05-0.2 -
1050°C 1oso”c 1050°C 1050°C 1050”c _ _
tions formed during irradiation influence their mobility, capability for point defect absorption and dynamical balance of point defects. The main characteristics of dislocation structure (concentration, size, value and geometry of Burgers vector) determine the intensity of radiation-induced solute segregation. For all the irradiated stainless steels investigated by us a general feature is the presence of the stage of formation and growth of faulted Frank loops. The loop growth is realized through interstitial atom absorption on the (111) planes along the lines (110). The last is confirmed by the loop edge elongation along the (110) direction. Evaluation of the interstitial absorption rate in this direction gives a value of N 10’ nm/dpa under irradiation of 16Cr-15Ni steel in the high voltage electron microscope (HVEM). At some dose level (lo-15 dpa) the faulted loops are transformed into the unfaulted ones according to the known reaction [3]: (a/3)(111)
+ (a/6)(112)
= (a/2)(110>.
The loss of the loop faultness in austenitic steels, especially in the deformed state, is possibly also due to the interaction between Frank loops with Burgers vector b = (a/3)( 111) and twinning dislocations with Burgers vector b = (a/6)( 112) in accordance with the same reaction [4]. The presence of interstitial gas and other solute atoms in the solid solution leads to the reduction in sizes and increase in concentration of dislocation loops. It is interesting to note that dislocation loops arising at dose levels higher than lo-15 dpa are perfect as a rule. In the cold worked (CW) steels the faultiness Table 2 Irradiation
of Nuclear Materials 212-215 (1994) 39-44
disappears under higher dose levels (lo-15 dpa). For aged specimens (8OO”C,200 h) the stage of faulted loop existence shrinks sharply. The loss of faultiness under certain irradiation doses, the nucleation of unfaulted loops under high irradiation doses, the influence of cold working and thermomechanical treatment on the extent of the faulted loop regime allow conclusions to be drawn about the increase in stacking fault energy with the dose. This fact may be explained by the changes in solute and component composition under irradiation. In the austenitic steels with a relatively stable phase composition under irradiation, the dislocation structure evolution plays a particular role in the development of vacancy voidage. It is necessary to maintain a high dislocation density in order to increase the incubation period. One of the methods for dislocation structure stabilization is an alloying with the elements which segregate on dislocations and thus decrease their climb rate. Such solutes as Si, Ti, B and rare earth elements have an influence on all components of the dislocation structure arising during irradiation. The boron addition to 16Cr-lSNi-3Mo-Nb steel moderates the dislocation structure evolution, extending the faulted Frank loop stage. It decreases stacking fault energy and reduces elastic stresses around dislocations. All these factors promote an increase in the incubation period. Unlike austenitic steels, irradiation in the ferriticmartensitic steels 13Cr-2Mo-NbVB, alloys Fe + 13Cr, 13Cr-2Mo + TiO, leads to formation of mainly perfect prismatic loops with Burgers vectors [loo] ([S]) on the (100) planes. The processes of nucleation and
conditions
Irradiation
Energy
Defect
Irradiation
Total dose
device
particles
production
temperature PC)
(dpa)
(MeV)
rate (dpa/s)
5 1
l-3 x 10-l 1-3x 10-l 10-5-10-6
350-800 270-700 300-650
l-200 0.1-100 5 -80
lo-s-lo-6
400-650
10-75
ESUVI UTI-1 BOR-60 BN-600
> 0.1 > 0.1
of
I.M. Neklyudov, V.N. Voyevodin/Journal of Nuclear Materials212-215 (1994) 39-44
growth of interstitial loops in 13Cr-2Mo-NbVB, Fe12Cr and 13Cr-2Mo + TiO, develop much more slowly than in the austenitic steels, thus promoting in a certain manner their radiation resistance increase. In the Fe-12Cr alloy the phenomenon of punching of the loops (previously described in Ref. [6]) was found together with ordinary loop nucleation. The formation of the punched loop groups takes place on the (100) slip plane along the (100) direction. The theory of this process was described elsewhere. The formation of ordered loops on the (100) and (110) planes is a characteristic feature of the 13Cr-2Mo + TiO, steel and the alloys based on Cr. Most of them ( * 90%) have Burgers vector (100) and may be related to radiation-induced segregation [S]. The relative “friability” of the bee lattice and the associated high diffusional mobility of interstitial elements (C and N) lead to the fact that microstructural evolution during irradiation of ferritic alloys is determined mainly by the behaviour of the interstitial elements. It is evident from Fig. 1 that the interstitial elements affect the beginning of dislocation loop nucleation, their size and concentration significantly. A high bonding energy between interstitial elements promotes an increase in the number of loop nucleation points, while segregation of nitrogen atoms on the loops impedes their growth. The same slope on the plots of dependencies of loop size with dose in alloys with different nitrogen concentrations testifies to the slight influence of nitrogen on the loop growth rate. 3.2. Segregation phenomena An intensive segregation of atoms to neutral and preferential sinks is determined by the diffusion pa-
16
m A F.-12%Cr+O.002N
14 -
0 A Fe-12%CH O.lN
.
0
I 0
2
I 4
I 6 Dose,
I 8
I
I
10
12
lE+l6 14
d.p.a.
Evolution of size and number density of dislocation loops in Fe-12Cr alloys with different nitrogen content (Cr3+, E = 1 MeV, Ti,, = 4OO”C,D = 9 dpa).
Fig. 1.
ocr-1oNl.n. wN-60.
l,=eat,
046 rIpa.
.“--\r----MO
-240
-100
-120
40
0
60
120
130
240
Jo0
Distance from boundary, nm
Fig. 2. Segregation profile (migrating boundary, 18CrlONiTi, BOR-60,
T,,, = 5WC,
D = 53 dpa).
rameters of matrix atoms for vacancy and interstitial mechanisms [7]. These two mechanisms can compete with one another or act cooperatively. The decay of solid solution is dependent on interrelation of the fluxes functioning by means of the different mechanisms determined by partial diffusion coefficients which are very sensitive to composition and alloy structure. The segregation profiles for neutral sinks in the form of stationary boundaries of austenite are characterized by the symmetrical enrichment in undersized Ni and Si, and depletion in oversized atoms of Cr and MO. Grain boundary migration at a high irradiation temperature breaks this symmetry: regions depleted in Ni and enriched in Cr arise. The behaviour of the oversized Ti atom turned out to be obscure: an enrichment in Ti on boundaries was observed although the solutes with large atomic radii usually deplete boundaries (Fig. 2). The data for temperature dependence of segregation on boundaries indicate a difference in vacancy behaviour under irradiation temperature change reflecting the ability for recombination and formation of point defect flows necessary for segregation. An important conclusion from the examinations performed for 13Cr-2Mo-NbVB steel is the depletion of boundaries in Cr during irradiation. Cr as an oversized atom for the Fe-Cr system (d _ 4.36) segregates from a boundary by means of vacancy flow in accordance with the inverse Kirkendall effect. The formation of the “shell” consisting of Ni and Si atoms around voids in irradiated steel 18Cr-lONi-Ti can also be explained by the manifestation of the inverse Kirkendall effect [8]. For two-phase steel 26Cr-6Ni we have measured the distribution of its main components (Fe, Cr, Ni, Si) over the cross section of the grains making up the mutual boundary, namely: “austenite/ austenite”, “austenite/ ferrite” and “ferrite/ ferrite” before and
42
I.M. Neklyudoc, V.N. Voyecbodin/Journal of Nuclear Marerials 212-215 (I9941 39-44
after irradiation with Cr ions. It was established that the supposed element distribution is modified during irradiation, and the degree of modification is dependent on boundary type. The boundaries between one phase grains (“austenite/ austenite” and “ferrite/ ferrite”) are practically symmetrical: on the boundaries “austenite/austenite” a depletion in Cr and Fe takes place, but there is a tendency toward an increase in Ni and Si contents. The boundaries “ferrite/ferrite” are characterized during irradiation by the larger degree of depletion in Cr and the growth of Ni, Si and Fe concentrations. The natural asymmetry in the segregation profile distribution for the “austenite/ferrite” boundary in the initial condition is characterized during irradiation by an increase in the misfit in element quantity in comparison with other boundary types. The preferential sinks (dislocation loops) in the investigated austenitic steels are characterized mainly (concerning the principal element segregation) by the same tendencies as those described for neutral ones. The enrichment of loops in Ni, Si (although in a smaller degree) and depletion in Cr and Fe take place. The average relations of enrichment to depletion are: 1.02, 1.03, 0.98, and 0.96 respectively. These results coincide with data obtained on the US PCA steel irradiated in FFTF [9]. As noted above, in the ferritic steels perfect loops having small sizes and high concentration were observed. Typical segregation profiles for dislocation loops with Burgers vector ~(100) in the steel 13Cr2Mo-NbVB irradiated with chromium ions at 575°C are shown in Fig. 3. A confirmation of Cr segregation on the perfect dislocation loops with Burgers vector ~(100) is the fact that on the ordered (formed up in rows) dislocation
loops on the (100) planes a formation second phase precipitates M,X takes place with increase of dose. In that phase the M is mainly Cr according to energy-dispersive X-ray analysis. It is important that this phase precipitation occurs on the ~(100) type dislocation loops since the habitus plane for the M,X phase coincides with the plane of
of orientated
the given type loops (loo),. On the (a/2)(111) type loops lying on the (loo), planes, the segregation of interstitial elements occurs to a significantly lesser degree. The results presented confirm the close interdependence between dislocation structure evolution and seg-
16
5
16 s * t
14
I 0
I
I
I
I
I
I
I
l
l
10
20
30
40
60
60
70
60
60
0 100
Distance from dislocation, nm
Fig. 3. Segregation profiles (dislocation loop in 13Cr2MoNbVB, Cr3+, E = 1 MeV, Tin = 575”C, D = 48 dpa).
processes. The degree of segregation depends on the type of dislocation structure in irradiated material. The results on enrichment of dislocation loops with Cr in the ferritic steels are in good agreement with the data in Ref. [lo]. It is shown in Fig. 3 that the changes in Si and MO content in the loops of the ferritic steel are the same as in the loops of austenitic steels, namely enrichment in Si and depletion in MO. regation
3.3. Phase transformations The acceleration of nucleation and growth processes for equilibrium phases, dissolution of thermally stable phases and formation of nonequilibrium (radiation-induced) ones are observed in materials during irradiation. This work presents the results of an examination of second phase precipitate stability at irradiation temperatures for which the main mechanism of the phase stability loss is a radiation-induced segregation. Particular attention was paid to the phases having a large degree of misfit (nonisomorphic) with the matrix lattice. They are: phosphides, carbonitrides, hphase in austenitic steels and Laves in ferritic steels. The evolution of element composition and structural state of the phosphides during irradiation was studied for the specimens of radiation-resistant austenitic steel 16Cr-llNi-3Mo-Ti alloyed with SC [12]. In nonirradiated specimens of this steel (Ni, SC&P
Table 3 Change of phosphides composition (wt%) Unirradiated 40 dpa 100 dpa
Fe
Cr
Ni
Mn
P
Si
SC
9.3 11.2 12.7
3.1 1.4 3.3
45.1 41.9 50.3
2.0
11.3 8.1 2.9
4.2 6.2
31.2 28.9 12.6
2.2
(NiScl2P (NiSc12P G-phase
I.&l. Ne~~ov,
f/.N. Voyevodin/.Joumal of Nuciear
Ma#e~ls
212-215
(1994)
43
39-44
Table 4 Change of MC composition (wt%) Uni~adiated 50 dpa 150 dpa
NIJ
MO
Fe
Cr
Ni
Mn
Si
99 72 11
0.5 2 3
0.3 10 16
0.2 3 4
9 51
3
4 12
phosphides together with MX precipitates are observed. They have a hexagonal crystal lattice with parameters a = 0.59 nm and c = 0.34 nm. Irradiation with Cr ions at 600°C up to 100 dpa leads to changes and transformation into G-phase (Table 3). In specimens of this steel irradiated up to 70 dpa in BOR-60 both phosphides and large G-phase precipitates containing about 7 wt% of phosphorus were observed. During the first stage of irradiation of austenitic steels stabilized with Ti and Nb, finely dispersed carbonit~des of these elements precipitate as type MX. Having a large misfit parameter, MX phases keep their coherence only under very small sizes. With increasing irradiation dose the precipitates grow in size and, at a definite stage, the precipitates lose their coherence with the matrix. As a rest&, their dissolution occurs. During irradiation the composition of carbonitrides is modified, being enriched in Ni and Si (Table 4). Gphase was a principal one precipitated under irradiation in BOR-60 up to 68 dpa at temperatures ranging from 360 to 6OO”C,whereas at a dose of 11 dpa a fine (3-5 nm) Ti monocarbides and primary MX type carbides enriched in nickel and silicon were found. The change in phase composition with dose increase is explained by the infiltration of Ni and Si into MX
Crl3N:13HdT,3
9
635 c.
I
3 ilsv Cd’
A Ni$i-pLatelets . Ni~l~-gto6utes
: :
Fig. 4. Microstructural development in Ti-modified Cr13Ni13 austenitic stainless steel irradiated with 3 MeV Cr3+ ions at 635°C (dose rate - lo-’ dpa/sf. At doses greater than - 130 dpa semicoherent platelets (Ni,Ti) transform into incoherent globular particles.
MC MC G-phase
phases with incoherent interphases and MX-G transition where (in G-phase) interstitial elements are dissolved. Under irradiation of 13Cr-13Ni-3Ti steel the precipitation of thin platelet hcp h-phase (Ni,Ti) takes place. The size of these precipitates grows with irradiation dose. The precipitates conjugate coherently with the matrix basic plane and incoherently in planes perpendicular to the basic one. At the expense of this the fluxes of point defects and hence solute atoms to incoherent interfaces of h-phase precipitate significantly exceed those to coherent interfaces, which is the cause of the precipitate growth as thin platelets. At dose levels more than 130 dpa the Ni,Ti precipitates change their form to a globular one (Fig. 4) which is completely incoherent. The consequence of this change is a transformation of h-phase globular precipitates to G-phase ones with a more coherent precipitate-matrix interface. In lOCr-6Mo-NbV ferritic steel, after annealing at 730°C for 1 h, Laves phase was formed and was found to be distributed homogeneously in grain bodies. After irradiation with Cr ions at 450°C up to 150 dpa, dissolution of these phases took place. The dissolution rate of Laves increases with irradiation temperature. At the temperature of 500°C and dose of 150 dpa a complete dissolution of Fe,(Mo, Nbl takes place and only h-phase precipitates having a coherent conjugation with the matrix were observed. Thus, in spite of differences in radiation-induced segregation processes for austenitic and ferritic materials, the behaviour of precipitates during irradiation of these steels has much in common. The radiation-induced segregation of solute elements to interfaces leads to alloy composition change in the precipitate environment and hence to changes in their stability. The most stable precipitates under irradiation are those having a good c~stallographic conjugation with the matrix. The composition of incoherent precipitates is changed during irradiation. With increasing irradiation dose the incoherent precipitates either dissolve or transform into precipitates with a low degree of incoherence.
4. Conclusions The principal cause of structure-phase transformations in steels and alloys is a high concentration of
44
I.M. Neklyudou, VN. Voyeuodin /Journal
point defects, the presence of neutral and preferential sinks and occurrence of point defect directed fluxes. The dislocation structure evolution rate is higher in austenitic steels than in ferritic ones. The stage of faulted Frank loops observed in the austenitic steels is almost absent in the ferritics. The moment of loss of the loop faultiness is defined by an increase in stacking fault energy because of the change in material composition during irradiation. The difference in solid solution decay is determined by the behaviour of principal segregants; in austenitic steels neutral and preferential sinks are enriched in Ni and Si and depleted in Cr, MO, Nb; whilst in the ferritic steels, neutral sinks (on the whole-grain boundaries) are depleted in Cr, but preferential sinks are enriched in Cr, and, in addition, during irradiation a selected enrichment in Cr of dislocation loops of type a( 100) lying on the {loo) planes takes place. The radiation-induced segregation of alloy components and alloying elements to interphase boundaries leads to precipitate composition change and, consequently to alternation of their stability. The most stable
precipitates during irradiation are those having a good crystallographic conjugation with the matrix. With irradiation dose increase, an incoherent precipitate is either dissolved or transformed into one with a lesser degree of incoherence. The composition of second phase precipitates is determined mainly by partial diffusion coefftcients of the elements in solid solution.
of Nuclear Materials 212-215
(1994) 39-44
Alloying with oversized and undersized elements changes the stability of the dislocation structure and influences point defect recombination degree and solid solution decay processes.
References
[ll V.F. Zelensky, I.M. Neklyudov, V.N. Voyevodin, Phyzika Chimiya Obrabotki Materialov 4 (1991).
Dl D.R. Harries, Proc. Consult. Symp. Harwell, September 9-11, 1974, Rep. AERE-R-7934, (1975) p. 287. t31 D.S. Gelles, Effects of Radiation on Materials, 12th Int.
Symp., ASTM-SIP 870, eds. F.A. Garner and J.S. Perrin, (ASTM, Philadelphia, 1985) pp. 98- 112. 141 V.N. Voyevodin, I.M. Neklyudov and P.V. Platonov, Voprosy Atomnoy Nauki Tehniki: FRP i RM, 3c.50)(1989). [51 V.N. Voyevodin, V.F. Zelensky, I.M. Neklyudov, Effects of Radiation on Materials, 15th Int. Symp., ASTM-STP 1125 (1992). 161O.V. Borodin, V.N. Voyevodin and I.M. Neklyudov, Voprosy Atomnoy Nauki i Tehniki: FRP RM 3(50) (1989). [71 V.F. Zelensky, I.M. Neklyudov and V.N. Voyevodin, Proc. 14th Int. Symp. on Effects of Radiation on Materials, Andover, USA, 1988, p. 193. Ml AS. Bakai, et al., J. Nucl. Mater. 185 (1991) 260. [91 E.A. Kenik, Scripta Metall. 10 (1976) 733. [lOI N. Yoshida, A. Yamaguchi and T. Muroga, J. Nucl. Mater. 155-157 (1988) 1232. [ill V.V. Bryk, I.M. Neklyudov and P.V. Platonov, Mater. Sci. Forum 97-99 (1992).
Journal of Nuclear Materials 212-21.5 (1994) 39-44
Features of structure-phase transformations and segregation processes under irradiation of austenitic and ferritic-martensitic steels I.M. Neklyudov, V.N. Voyevodin Kharkov Institute of Physics and Technology, 310108 Kharkou, Ukraine
Abstract The difference between crystal lattices of austenitic and ferritic steels leads to distinctive features in mechanisms of physical-mechanical change. This paper presents the results of investigations of dislocation structure and phase evolution, and segregation phenomena in austenitic and ferritic-martensitic steels and alloys during irradiation with heavy ions in the ESUVI and UT1 accelerators and by neutrons in fast reactors BOR-60 and BN-600. The influence of different factors (including different alloying elements) on processes of structure-phase transformation was studied.
1. Introduction
2. Experimental procedures and materials
Complex interdependent processes take place in solids during their irradiation: nuclear processes resulting in appearance of primary knockon atoms and new elements; atom-atomic cascade collisions, which produce crystal lattice imperfections in the form of point defects, crowdions and depleted zones; substructural processes leading to the formation of directly registered clusters, nuclei of dislocation loops, precipitate particles and voids; diffusional ones which are responsible for evolution of micro- and macrostructures, solute and phase compositions. The substructural and diffusional processes play a major role in physicalmechanical property changes. Stainless austenitic and ferritic-martensitic steels are candidates for structural materials for fusion reactor core applications. The difference in their chemical composition and crystal structure results in distinctive features of the diffusional and substructural processes occurring during irradiation and, consequently, in the radiation resistance of these steels [1,2]. This paper is a brief review of results on some aspects of structure-phase transformations and radiation-induced segregation mechanisms under irradiation of austenitic and ferritic steels.
The chemical composition and thermal treatment of the steels studied are given in Tables 1 and 2. The structure-phase state of initial and irradiated steel specimens was examined with the aid of transmission electron microscopes (TEMs): EM-125K TEM and JEM-IOOCX TEM equipped with ASID-4D adapter and X-ray energy-dispersive spectrometer Link Systems-860. A quantitative element analysis was carried out using the RTS2/FLS program. For irradiations the ESUVI (material research heavy ion accelerator), the UT1 (heavy ion accelerator), and fast reactors BOR-60 and BN-600 were used, with the irradiation conditions given in Table 2.
3. Results and discussion 3.1. Evolution
of dislocation
structure
The radiation resistance of steels is determined mainly by the rate of nucleation and evolution of dislocation loops. The type and parameters of disloca-
0022-3115/94/$07.00 0 1994 Elsevier Science B.V. All rights reserved SSDI 0022-3115(94)00277-U
I.M. Neklyudor~, V.N. Vp~~erndin/Journal
40 Table 1 Chemical composition
18Cr-1ONi 16Cr-15Ni 13Cr-13Ni 16Cr-11Ni 13Cr-2Mo 9Cr-2Mo 26Cr-6Ni
of investigated
materials
(wt%)
Ni
Cr
Mn
MO
C
Ti
Nb
Si
SC
Thermal
10.6 14.5 13.2 10.7 0.2 _ 5.46
18.0 15.5 13.1 16.2 12.1 9.42 25.6
1.2 0.8 2.5 1.3 0.5 0.7 0.8
2.9 0.5 2.3 1.8 1.5 -
0.04 0.04 0.06 0.15 0.13 0.02
0.06 2.7 -
0.8 _ _ 0.25 0.31 -
0.8 0.3 _ 0.6 0.5 0.6 0.6
0.02-0.05 0.05-0.2 0.05-0.2 -
1050°C 1oso”c 1050°C 1050°C 1050”c _ _
tions formed during irradiation influence their mobility, capability for point defect absorption and dynamical balance of point defects. The main characteristics of dislocation structure (concentration, size, value and geometry of Burgers vector) determine the intensity of radiation-induced solute segregation. For all the irradiated stainless steels investigated by us a general feature is the presence of the stage of formation and growth of faulted Frank loops. The loop growth is realized through interstitial atom absorption on the (111) planes along the lines (110). The last is confirmed by the loop edge elongation along the (110) direction. Evaluation of the interstitial absorption rate in this direction gives a value of N 10’ nm/dpa under irradiation of 16Cr-15Ni steel in the high voltage electron microscope (HVEM). At some dose level (lo-15 dpa) the faulted loops are transformed into the unfaulted ones according to the known reaction [3]: (a/3)(111)
+ (a/6)(112)
= (a/2)(110>.
The loss of the loop faultness in austenitic steels, especially in the deformed state, is possibly also due to the interaction between Frank loops with Burgers vector b = (a/3)( 111) and twinning dislocations with Burgers vector b = (a/6)( 112) in accordance with the same reaction [4]. The presence of interstitial gas and other solute atoms in the solid solution leads to the reduction in sizes and increase in concentration of dislocation loops. It is interesting to note that dislocation loops arising at dose levels higher than lo-15 dpa are perfect as a rule. In the cold worked (CW) steels the faultiness Table 2 Irradiation
of Nuclear Materials 212-215 (1994) 39-44
disappears under higher dose levels (lo-15 dpa). For aged specimens (8OO”C,200 h) the stage of faulted loop existence shrinks sharply. The loss of faultiness under certain irradiation doses, the nucleation of unfaulted loops under high irradiation doses, the influence of cold working and thermomechanical treatment on the extent of the faulted loop regime allow conclusions to be drawn about the increase in stacking fault energy with the dose. This fact may be explained by the changes in solute and component composition under irradiation. In the austenitic steels with a relatively stable phase composition under irradiation, the dislocation structure evolution plays a particular role in the development of vacancy voidage. It is necessary to maintain a high dislocation density in order to increase the incubation period. One of the methods for dislocation structure stabilization is an alloying with the elements which segregate on dislocations and thus decrease their climb rate. Such solutes as Si, Ti, B and rare earth elements have an influence on all components of the dislocation structure arising during irradiation. The boron addition to 16Cr-lSNi-3Mo-Nb steel moderates the dislocation structure evolution, extending the faulted Frank loop stage. It decreases stacking fault energy and reduces elastic stresses around dislocations. All these factors promote an increase in the incubation period. Unlike austenitic steels, irradiation in the ferriticmartensitic steels 13Cr-2Mo-NbVB, alloys Fe + 13Cr, 13Cr-2Mo + TiO, leads to formation of mainly perfect prismatic loops with Burgers vectors [loo] ([S]) on the (100) planes. The processes of nucleation and
conditions
Irradiation
Energy
Defect
Irradiation
Total dose
device
particles
production
temperature PC)
(dpa)
(MeV)
rate (dpa/s)
5 1
l-3 x 10-l 1-3x 10-l 10-5-10-6
350-800 270-700 300-650
l-200 0.1-100 5 -80
lo-s-lo-6
400-650
10-75
ESUVI UTI-1 BOR-60 BN-600
> 0.1 > 0.1
of
I.M. Neklyudov, V.N. Voyevodin/Journal of Nuclear Materials212-215 (1994) 39-44
growth of interstitial loops in 13Cr-2Mo-NbVB, Fe12Cr and 13Cr-2Mo + TiO, develop much more slowly than in the austenitic steels, thus promoting in a certain manner their radiation resistance increase. In the Fe-12Cr alloy the phenomenon of punching of the loops (previously described in Ref. [6]) was found together with ordinary loop nucleation. The formation of the punched loop groups takes place on the (100) slip plane along the (100) direction. The theory of this process was described elsewhere. The formation of ordered loops on the (100) and (110) planes is a characteristic feature of the 13Cr-2Mo + TiO, steel and the alloys based on Cr. Most of them ( * 90%) have Burgers vector (100) and may be related to radiation-induced segregation [S]. The relative “friability” of the bee lattice and the associated high diffusional mobility of interstitial elements (C and N) lead to the fact that microstructural evolution during irradiation of ferritic alloys is determined mainly by the behaviour of the interstitial elements. It is evident from Fig. 1 that the interstitial elements affect the beginning of dislocation loop nucleation, their size and concentration significantly. A high bonding energy between interstitial elements promotes an increase in the number of loop nucleation points, while segregation of nitrogen atoms on the loops impedes their growth. The same slope on the plots of dependencies of loop size with dose in alloys with different nitrogen concentrations testifies to the slight influence of nitrogen on the loop growth rate. 3.2. Segregation phenomena An intensive segregation of atoms to neutral and preferential sinks is determined by the diffusion pa-
16
m A F.-12%Cr+O.002N
14 -
0 A Fe-12%CH O.lN
.
0
I 0
2
I 4
I 6 Dose,
I 8
I
I
10
12
lE+l6 14
d.p.a.
Evolution of size and number density of dislocation loops in Fe-12Cr alloys with different nitrogen content (Cr3+, E = 1 MeV, Ti,, = 4OO”C,D = 9 dpa).
Fig. 1.
ocr-1oNl.n. wN-60.
l,=eat,
046 rIpa.
.“--\r----MO
-240
-100
-120
40
0
60
120
130
240
Jo0
Distance from boundary, nm
Fig. 2. Segregation profile (migrating boundary, 18CrlONiTi, BOR-60,
T,,, = 5WC,
D = 53 dpa).
rameters of matrix atoms for vacancy and interstitial mechanisms [7]. These two mechanisms can compete with one another or act cooperatively. The decay of solid solution is dependent on interrelation of the fluxes functioning by means of the different mechanisms determined by partial diffusion coefficients which are very sensitive to composition and alloy structure. The segregation profiles for neutral sinks in the form of stationary boundaries of austenite are characterized by the symmetrical enrichment in undersized Ni and Si, and depletion in oversized atoms of Cr and MO. Grain boundary migration at a high irradiation temperature breaks this symmetry: regions depleted in Ni and enriched in Cr arise. The behaviour of the oversized Ti atom turned out to be obscure: an enrichment in Ti on boundaries was observed although the solutes with large atomic radii usually deplete boundaries (Fig. 2). The data for temperature dependence of segregation on boundaries indicate a difference in vacancy behaviour under irradiation temperature change reflecting the ability for recombination and formation of point defect flows necessary for segregation. An important conclusion from the examinations performed for 13Cr-2Mo-NbVB steel is the depletion of boundaries in Cr during irradiation. Cr as an oversized atom for the Fe-Cr system (d _ 4.36) segregates from a boundary by means of vacancy flow in accordance with the inverse Kirkendall effect. The formation of the “shell” consisting of Ni and Si atoms around voids in irradiated steel 18Cr-lONi-Ti can also be explained by the manifestation of the inverse Kirkendall effect [8]. For two-phase steel 26Cr-6Ni we have measured the distribution of its main components (Fe, Cr, Ni, Si) over the cross section of the grains making up the mutual boundary, namely: “austenite/ austenite”, “austenite/ ferrite” and “ferrite/ ferrite” before and
42
I.M. Neklyudoc, V.N. Voyecbodin/Journal of Nuclear Marerials 212-215 (I9941 39-44
after irradiation with Cr ions. It was established that the supposed element distribution is modified during irradiation, and the degree of modification is dependent on boundary type. The boundaries between one phase grains (“austenite/ austenite” and “ferrite/ ferrite”) are practically symmetrical: on the boundaries “austenite/austenite” a depletion in Cr and Fe takes place, but there is a tendency toward an increase in Ni and Si contents. The boundaries “ferrite/ferrite” are characterized during irradiation by the larger degree of depletion in Cr and the growth of Ni, Si and Fe concentrations. The natural asymmetry in the segregation profile distribution for the “austenite/ferrite” boundary in the initial condition is characterized during irradiation by an increase in the misfit in element quantity in comparison with other boundary types. The preferential sinks (dislocation loops) in the investigated austenitic steels are characterized mainly (concerning the principal element segregation) by the same tendencies as those described for neutral ones. The enrichment of loops in Ni, Si (although in a smaller degree) and depletion in Cr and Fe take place. The average relations of enrichment to depletion are: 1.02, 1.03, 0.98, and 0.96 respectively. These results coincide with data obtained on the US PCA steel irradiated in FFTF [9]. As noted above, in the ferritic steels perfect loops having small sizes and high concentration were observed. Typical segregation profiles for dislocation loops with Burgers vector ~(100) in the steel 13Cr2Mo-NbVB irradiated with chromium ions at 575°C are shown in Fig. 3. A confirmation of Cr segregation on the perfect dislocation loops with Burgers vector ~(100) is the fact that on the ordered (formed up in rows) dislocation
loops on the (100) planes a formation second phase precipitates M,X takes place with increase of dose. In that phase the M is mainly Cr according to energy-dispersive X-ray analysis. It is important that this phase precipitation occurs on the ~(100) type dislocation loops since the habitus plane for the M,X phase coincides with the plane of
of orientated
the given type loops (loo),. On the (a/2)(111) type loops lying on the (loo), planes, the segregation of interstitial elements occurs to a significantly lesser degree. The results presented confirm the close interdependence between dislocation structure evolution and seg-
16
5
16 s * t
14
I 0
I
I
I
I
I
I
I
l
l
10
20
30
40
60
60
70
60
60
0 100
Distance from dislocation, nm
Fig. 3. Segregation profiles (dislocation loop in 13Cr2MoNbVB, Cr3+, E = 1 MeV, Tin = 575”C, D = 48 dpa).
processes. The degree of segregation depends on the type of dislocation structure in irradiated material. The results on enrichment of dislocation loops with Cr in the ferritic steels are in good agreement with the data in Ref. [lo]. It is shown in Fig. 3 that the changes in Si and MO content in the loops of the ferritic steel are the same as in the loops of austenitic steels, namely enrichment in Si and depletion in MO. regation
3.3. Phase transformations The acceleration of nucleation and growth processes for equilibrium phases, dissolution of thermally stable phases and formation of nonequilibrium (radiation-induced) ones are observed in materials during irradiation. This work presents the results of an examination of second phase precipitate stability at irradiation temperatures for which the main mechanism of the phase stability loss is a radiation-induced segregation. Particular attention was paid to the phases having a large degree of misfit (nonisomorphic) with the matrix lattice. They are: phosphides, carbonitrides, hphase in austenitic steels and Laves in ferritic steels. The evolution of element composition and structural state of the phosphides during irradiation was studied for the specimens of radiation-resistant austenitic steel 16Cr-llNi-3Mo-Ti alloyed with SC [12]. In nonirradiated specimens of this steel (Ni, SC&P
Table 3 Change of phosphides composition (wt%) Unirradiated 40 dpa 100 dpa
Fe
Cr
Ni
Mn
P
Si
SC
9.3 11.2 12.7
3.1 1.4 3.3
45.1 41.9 50.3
2.0
11.3 8.1 2.9
4.2 6.2
31.2 28.9 12.6
2.2
(NiScl2P (NiSc12P G-phase
I.&l. Ne~~ov,
f/.N. Voyevodin/.Joumal of Nuciear
Ma#e~ls
212-215
(1994)
43
39-44
Table 4 Change of MC composition (wt%) Uni~adiated 50 dpa 150 dpa
NIJ
MO
Fe
Cr
Ni
Mn
Si
99 72 11
0.5 2 3
0.3 10 16
0.2 3 4
9 51
3
4 12
phosphides together with MX precipitates are observed. They have a hexagonal crystal lattice with parameters a = 0.59 nm and c = 0.34 nm. Irradiation with Cr ions at 600°C up to 100 dpa leads to changes and transformation into G-phase (Table 3). In specimens of this steel irradiated up to 70 dpa in BOR-60 both phosphides and large G-phase precipitates containing about 7 wt% of phosphorus were observed. During the first stage of irradiation of austenitic steels stabilized with Ti and Nb, finely dispersed carbonit~des of these elements precipitate as type MX. Having a large misfit parameter, MX phases keep their coherence only under very small sizes. With increasing irradiation dose the precipitates grow in size and, at a definite stage, the precipitates lose their coherence with the matrix. As a rest&, their dissolution occurs. During irradiation the composition of carbonitrides is modified, being enriched in Ni and Si (Table 4). Gphase was a principal one precipitated under irradiation in BOR-60 up to 68 dpa at temperatures ranging from 360 to 6OO”C,whereas at a dose of 11 dpa a fine (3-5 nm) Ti monocarbides and primary MX type carbides enriched in nickel and silicon were found. The change in phase composition with dose increase is explained by the infiltration of Ni and Si into MX
Crl3N:13HdT,3
9
635 c.
I
3 ilsv Cd’
A Ni$i-pLatelets . Ni~l~-gto6utes
: :
Fig. 4. Microstructural development in Ti-modified Cr13Ni13 austenitic stainless steel irradiated with 3 MeV Cr3+ ions at 635°C (dose rate - lo-’ dpa/sf. At doses greater than - 130 dpa semicoherent platelets (Ni,Ti) transform into incoherent globular particles.
MC MC G-phase
phases with incoherent interphases and MX-G transition where (in G-phase) interstitial elements are dissolved. Under irradiation of 13Cr-13Ni-3Ti steel the precipitation of thin platelet hcp h-phase (Ni,Ti) takes place. The size of these precipitates grows with irradiation dose. The precipitates conjugate coherently with the matrix basic plane and incoherently in planes perpendicular to the basic one. At the expense of this the fluxes of point defects and hence solute atoms to incoherent interfaces of h-phase precipitate significantly exceed those to coherent interfaces, which is the cause of the precipitate growth as thin platelets. At dose levels more than 130 dpa the Ni,Ti precipitates change their form to a globular one (Fig. 4) which is completely incoherent. The consequence of this change is a transformation of h-phase globular precipitates to G-phase ones with a more coherent precipitate-matrix interface. In lOCr-6Mo-NbV ferritic steel, after annealing at 730°C for 1 h, Laves phase was formed and was found to be distributed homogeneously in grain bodies. After irradiation with Cr ions at 450°C up to 150 dpa, dissolution of these phases took place. The dissolution rate of Laves increases with irradiation temperature. At the temperature of 500°C and dose of 150 dpa a complete dissolution of Fe,(Mo, Nbl takes place and only h-phase precipitates having a coherent conjugation with the matrix were observed. Thus, in spite of differences in radiation-induced segregation processes for austenitic and ferritic materials, the behaviour of precipitates during irradiation of these steels has much in common. The radiation-induced segregation of solute elements to interfaces leads to alloy composition change in the precipitate environment and hence to changes in their stability. The most stable precipitates under irradiation are those having a good c~stallographic conjugation with the matrix. The composition of incoherent precipitates is changed during irradiation. With increasing irradiation dose the incoherent precipitates either dissolve or transform into precipitates with a low degree of incoherence.
4. Conclusions The principal cause of structure-phase transformations in steels and alloys is a high concentration of
44
I.M. Neklyudou, VN. Voyeuodin /Journal
point defects, the presence of neutral and preferential sinks and occurrence of point defect directed fluxes. The dislocation structure evolution rate is higher in austenitic steels than in ferritic ones. The stage of faulted Frank loops observed in the austenitic steels is almost absent in the ferritics. The moment of loss of the loop faultiness is defined by an increase in stacking fault energy because of the change in material composition during irradiation. The difference in solid solution decay is determined by the behaviour of principal segregants; in austenitic steels neutral and preferential sinks are enriched in Ni and Si and depleted in Cr, MO, Nb; whilst in the ferritic steels, neutral sinks (on the whole-grain boundaries) are depleted in Cr, but preferential sinks are enriched in Cr, and, in addition, during irradiation a selected enrichment in Cr of dislocation loops of type a( 100) lying on the {loo) planes takes place. The radiation-induced segregation of alloy components and alloying elements to interphase boundaries leads to precipitate composition change and, consequently to alternation of their stability. The most stable
precipitates during irradiation are those having a good crystallographic conjugation with the matrix. With irradiation dose increase, an incoherent precipitate is either dissolved or transformed into one with a lesser degree of incoherence. The composition of second phase precipitates is determined mainly by partial diffusion coefftcients of the elements in solid solution.
of Nuclear Materials 212-215
(1994) 39-44
Alloying with oversized and undersized elements changes the stability of the dislocation structure and influences point defect recombination degree and solid solution decay processes.
References
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