Comprehensive nuclear materials 2 09 properties of austenitic steels for nuclear reactor applications Comprehensive nuclear materials 2 09 properties of austenitic steels for nuclear reactor applications Comprehensive nuclear materials 2 09 properties of austenitic steels for nuclear reactor applications Comprehensive nuclear materials 2 09 properties of austenitic steels for nuclear reactor applications Comprehensive nuclear materials 2 09 properties of austenitic steels for nuclear reactor applications
Trang 1Reactor Applications
P J Maziasz and J T Busby
Oak Ridge National Laboratory, Oak Ridge, TN, USA
Published by Elsevier Ltd.
2.09.2.4 Precipitation Behavior During Elevated Temperature Aging 273
2.09.3 Summary of How Properties Can Change During Irradiation 275 2.09.4 Some Examples of Advanced Alloys for FBR and ITER/Fusion Applications 279
Abbreviations
ASTM American Society for Testing and
Materials
bcc Body-centered cubic
BWR Boiling water reactor
D-T Deuterium–tritium (fusion)
DBTT Ductile-to-brittle transition temperature
FBR Fast-breeder reactor
fcc Face-centered cubic
GenIV Generation IV reactors
HFIR High Flux Isotope Reactor
IASCC Irradiation-assisted stress-corrosion
cracking
ITER International Magnetic Fusion
demonstration device, being constructed
in Cadarache, France
LWR Light water reactor
MFR Magnetic fusion reactor
NIMS National Institute for Materials Science
(Japan)
ORR Oak Ridge Research Reactor
PCA Prime candidate alloy
PWR Pressurized water reactor
R&D Research and development
RIS Radiation-induced solute segregation
SA Solution annealed
SCC Stress-corrosion cracking
SEM Scanning electron microscopy
TEM Transmission electron microscopy UTS Ultimate tensile strength
YS Yield strength
2.09.1 Introduction Austenitic stainless steels are a class of materials that are extremely important to conventional and advanced reactor technologies, as well as one of the most widely used kinds of engineering alloys They are austenitic Fe–Cr–Ni alloys with 15–20Cr, 8–15Ni, and the balance Fe, because they have a face-centered-cubic (fcc) close-packed crystal structure, which imparts most of their physical and mechanical properties They are steels because they contain dis-solved C, typically 0.03–0.15%, and more advanced steels can also contain similar or greater amounts
of dissolved N They are stainless because they con-tain>13%Cr and Cr provides surface passivation for corrosion-resistance in various aqueous or corrosive chemical environments from room temperature to about 400C At elevated temperatures of 500C and above, Cr provides oxidation resistance by the forma-tion of protective Cr2O3 oxide scales Commercial stainless steels are complex alloys, with varying addi-tions and combinaaddi-tions of Mo, Mn, Si, and Ti as well
as Nb to enhance the properties and behavior of the austenite parent phase over a wide range of
267
Trang 2temperatures They can also contain a host of minor or
impurity elements, including Co, Cu, V, P, B, and S,
which do not have significant effects within certain
normal ranges
Typical commercial steel grades relevant to
nuclear reactor applications include types 304, 316,
321, and 347 They can be fashioned into a wide
range of thick or thin components by hot or cold
rolling, bending, forging, or extrusion, and many are
also available as casting grades as well (i.e., 304 as
CF8, 316 as CF8M, and 347 as CF8C) These steels
all have good combinations of strength and ductility
at both high and low temperatures, with excellent
fatigue resistance, and are most often used in the
solution-annealed (SA) condition, with the alloying
elements fully dissolved in the parent austenite
phase and little or no precipitation The steels with
added Mo (316) or stabilized with Ti (321) or Nb
(347) also have reasonably good elevated temperature
strength and creep resistance Additions of nitrogen
(i.e., 316LN or 316N) provide higher strength and
stability of the austenite parent phase to the
embrit-tling effects of thermal- or strain-induced martensite
formation and allow this grade of steel to be used
at cryogenic temperatures It is beyond the scope
of this chapter to describe in detail the physical
metallurgy of austenitic stainless steels, and adequate
descriptions are found elsewhere.1,2 The remainder
of this chapter focuses on the factors that broadly
affect the properties of austenitic stainless steels in
specific reactor environments, and highlights efforts
to develop modified steels that perform significantly
better in such reactor systems These will likely be
important in enabling materials for any new
applica-tions of nuclear power
2.09.2 Properties of Unirradiated
Alloys
2.09.2.1 General and Fabrication Behavior
Without the effects of irradiation, austenitic stainless
steels are fairly stable solid-solution alloys that
generally remain in the metallurgical condition in
which they were processed at room temperature to
about 550C The typical austenitic stainless steel,
such as type 304, 316, 316L, or 347 stainless steel,
in the SA condition (1000–1050C), will have a
wrought, recrystallized grain structure of uniform,
equiaxed grains that are 50–100 mm in diameter,
particularly in products such as extruded bar or
flat-rolled plates (6–25 mm thick).1–3 Ideally, such
products should be free of plastic strain effects and have dislocation-free grains, but for real applications, products may be straightened or bent slightly (1–5% cold strain), and thus have some dislocation substruc-ture within the grains Stainless steel products with heavier wall thicknesses (>50 mm) would be forgings and castings, which would have coarser grain sizes, but probably not have additional deformation Spe-cial stainless steel products would include thin foils, sheets, or wires (0.08–0.5 mm thick), which would have much finer grain-sizes (1–10 mm diameter) due
to special processing (very short annealing times) and special considerations (5–10 grains across the foil/ sheet thickness).3Typical fast-breeder reactor (FBR) cladding for fuel elements can be thin-walled tubes
of austenitic stainless steel, with about 0.25 mm wall thickness, so they fall into this latter special pro-ducts category Although austenitic stainless steels are highly weldable, welding changes their structure and properties in the fusion (welded and resolidified) and adjacent heat-affected zones relative to the wrought base metal, so they may behave quite dif-ferently than the base metal, which is what was described above The detailed behavior of welds under irradiation is beyond the scope of this chapter,
so the remainder of this chapter focuses on typical wrought metal behavior
Another important aspect of austenitic stainless steel that defines it is the stability of the parent austenite phase The addition of nickel and elements that behave like nickel including carbon and nitrogen
to the alloy causes it to have the austenite parent phase and its beneficial properties, which is also the same fcc crystal structure found in nickel-based alloys Otherwise, the steel alloy would have the natural crystal structure of iron and chromium, which is body-centered cubic (bcc) ferrite, as the parent phase, and alloying elements that make the alloy behavior like this include molybdenum, nio-bium, titanium, vanadium, and silicon A stable aus-tenitic alloy will be 100% austenite, with no d-ferrite formed at high temperature and no thermal or strain-induced martensite, whereas an unstable austenitic alloy may have all of these A useful way of expressing these different phase formation tendencies at room temperature in terms of the alloy behaving more like
Cr (bcc ferritic) or Ni (fcc austenitic) is a Schaeffler diagram, as shown in Figure 1 The fcc austenite phase is nonmagnetic and maintains good strength and ductility even at cryogenic temperatures, with no embrittling effects of martensite formation The bcc phase by comparison is ferromagnetic, has a little less
Trang 3ductility (less active slip systems), and has a
ductile-to-brittle transition temperature (DBTT), below
which the steel has low ductility and impact
resis-tance, with a brittle fracture mode Maintaining
suf-ficient carbon and adding nitrogen are two ways of
imparting good, stable austenite phase behavior to
the common grades of austenitic stainless steels, like
304LN or 316LN
2.09.2.2 Physical Properties
Physical properties of 300 series stainless steels tend
to be fairly similar, and the typical physical
prop-erties of 316L stainless steel are given in Tables 1
and 2.1–3 The 316L stainless steel has a density
at room temperature of 8000 kg m3 and a melting
temperature of slightly above 1400C (Table 1)
The elastic (Young’s) modulus at room
tempera-ture is 190–200 GPa, which is typical of a range of
engineering alloys, including ferritic steels and
solid-solution Ni-based superalloys At 100C, the
coefficient of thermal expansion of 316L is about
16 106cm cm1C1 (Table 2), and values of
that property may vary by up to 3–4% for types
316 and 347 steels The 300 series stainless steels
have much more thermal expansion than martensi-tic/ferritic steels or Ni-based superalloys, with the thermal expansion of 316L at 100C being about
Table 1 Basic physical properties for 316L stainless steel
32
30
28
Austenite (A) Ferrite, 5%
Ferrite, 10%
20%
40%
80%
A + F
o
A + M + F
M + F
F + M
Martensite (M)
100%
ferrite Ferrite, 0%
Ferrite (F)
26
24
22
20
18
16
14
12
10
8
6
4
2
0
0 2 4 6 8 10 12
Chromium equivalent, %Cr + %Mo + 1.5 (%Si) + 0.5 (%Nb)
14 16 18 20 22 24 26 28 30 32 34 36 38 40 42
Figure 1 Schaeffler diagram showing regions of stable austenite, martensite, and delta-ferrite in austenitic stainless steels
at room temperature as a function of steel alloys compositional effects acting as the equivalent of Cr or Ni Reproduced from Lula, R A., Ed Stainless Steel; ASM International: Materials Park, OH, 1986.
Table 2 Thermal properties for 316L stainless steel Property Temperature
range
Value
Coefficient of thermal expansion
0–100C 15.9 10 6 C1 0–315C 16.2 10 6 C1 0–538C 17.5 10 6 C1 0–1000C 19.5 10 6 C1 Thermal
conductivity
At 100C 16.3 W mK1
At 500C 21.5 W mK1 Specific heat
capacity
0–100C 500 J kg1C
Trang 450% higher than that of type 410 ferritic steel.3The
thermal conductivity of 316L stainless steel at 100C
is 16.3 W mK1, which is to the higher end of the
range for such alloys, with type 316 or 347 steel having
15–30% lower thermal conductivity Thermal
con-ductivity of 300 series stainless steels is lower than
that of ferritic steels or Ni-based superalloys If the
300 series stainless steel is fully (100%) austenitic,
such as 316 or 347, then it has no ferromagnetic
behavior, but if it contains ferromagnetic phases (like
delta-ferrite or martensite), then such steels have some
degree of ferromagnetic behavior Adding nitrogen to
316L produces fully stable austenitic phase structures
2.09.2.3 Mechanical Properties
The general mechanical behavior properties of
austenitic stainless steels at room and at elevated
temperatures are described These provide the
back-ground for their behavior in various reactor
environ-ments The mechanical properties of the various
grades of the 300 series austenitic stainless steels are
fairly similar, particularly at room temperature, so
available data for type 316 or 316L steel are used as
representative of the group There is more variation
in properties at elevated temperatures, particularly
creep-resistance and creep–rupture strength, so
important properties differences are noted,
particu-larly for steels modified with Ti or Nb which have
more high-temperature heat-resistance than type
316 steel Some effects of processing on mechanical
properties are noted, but generally properties are
described for material in the SA condition
Austenitic stainless steels such as types 304, 316,
and 316L have yield strength (YS – 0.2% offset) of
260–300 MPa in the SA condition at room
tem-perature, with up to 50–70% total elongation.1–7
Typ-ical YS values as a function of temperature for type
316 are shown in Figures 2 and 3 Other austenitic
stainless steels developed for improved creep resistance
at high temperatures, such as fine-grained 347HFG
or the high-temperature, ultrafine
precipitate-strengthened (HT-UPS) steels (Table 3), have very
similar YS of about 250 MPa in the SA condition
(typical thicker section pipes or plates), as shown in
stainless steels require a minimum YS of 200 MPa
However, small amounts of cold plastic strain, 1–5%,
typical or straightening or flattening for various
prod-uct forms, termed ‘mill-annealed,’ raise the YS to
about 400 MPa, because austenitic stainless steels
tend to have high strain-hardening rates Large
amounts of cold work (CW) push the YS higher, with 20–30% CW 316 having YS of 600–700 MPa,8,9but with very low ductility of only 2–3% The very
450
350 300 250 200 150 100
50
Austenitic stainless steel
0
Figure 3 Comparison of yield strength (YS) at room temperature and at 700C for 316, 347HFG, and high-temperature, ultrafine precipitate-strengthened (HT-UPS) austenitic stainless steels, all in the solution-annealed condition, and for HT-UPS steel with 5%
CW prior to testing Adapted from Swindeman, R W.; Maziasz, P J.; Bolling, E.; King, J F Evaluation of Advanced Austenitic Alloys Relative to Alloy Design Criteria for Steam Service: Part 1 – Lean Stainless Steels; Oak Ridge National Laboratory Report (ORNL-6629/P1); Oak Ridge National Laboratory: Oak Ridge, TN, 1990; Teranishi, H.; et al.
In Second International Conference on Improved Coal Fired Power Plants; Electric Power Research Institute: Palo Alto,
CA, 1989; EPRI Publication GS-6422 (paper 33-1).
700
YLD
UTS
Temperature (⬚C)
600
500
400
300
200
100
0
Figure 2 Plots of yield strength (YS) and ultimate tensile strength (UTS) as a function of tensile test temperature for nine heats of SA 316 austenitic stainless steel tubing tested
by the National Research Institute for Metals (now NIMS) in Japan Reproduced from Data sheets on the elevated temperature properties of 18Cr–12Ni–Mo stainless steels for boiler and heat exchanger tubes (SUS 316 HTB), Creep Data Sheet No 6A; National Research Institute for Metals: Tokyo, Japan, 1978.
Trang 5fine grain sizes found in thin-sheet and foil products
made from 347 steel also tend to push ambient YS
to 275–300 MPa or above.7 The ultimate tensile
strength (UTS) of SA 316 steel at room temperature
is about 600 MPa, and can be higher (600–700 MPa) for
steels such as 347HFG, HT-UPS, or some of the
high-nitrogen grades The UTS of 20–30% CW 316 or other
comparable steels can be 700–800 MPa at room
temperature.4,8,9
The impact-toughness and crack-growth
resis-tance of SA 316 at room temperature and
tem-peratures below 500C are excellent because of its
high ductility and strain-hardening behavior Charpy
impact toughness values for SA 316 and 347 steel
are about 150 J at 22–400C, and tend to stay above
100 J even at cryogenic (196C) temperatures.
Type 316 stainless steels also have good
room-temperature fatigue resistance, exhibiting endurance
limits for cyclic stresses below the YS
At elevated temperatures, the YS of SA 316
declines with increasing temperature, reaching levels
of about 150 MPa at 600–650C (Figure 2), and
going lower at 700–800C More heat-resistant steels
such as 347HFG or HT-UPS steels may be slightly
stronger at 700C, and can have YS values of 300–
350 MPa in the ‘mill-annealed’ (5% CW) (Figure 3)
The UTS of SA 316 remains at about 500 MPa up to
500C, and then declines rapidly with increasing
temperature until YS and UTS approach similar
values (120–180 MPa) at about 800C (Figure 2)
More heat-resistant steel, such as 347HFG and the
HT-UPS steels, can retain higher UTS values of 200–
300 MPa at 800C Unaged SA 316 generally have
30–60% total tensile elongation at temperatures up
to 800C; similar steels with 20–30% CW can have
5–10% ductility until they recrystallize at
tempera-tures of 800C or above.9
At elevated temperatures, time-dependent defor-mation, or creep, becomes a concern for austenitic steels such as 304 and 316 above 500–550C
A Larson–Miller parameter (LMP) plot of creep– rupture strength for SA 316 is shown in Figure 4, and for 347HFG and HT-UPS steels in Figure 5 Long-term creep–rupture behavior is affected by precipitation behavior at elevated temperatures, as
is described in the following section Creep–rupture behavior (time to rupture or time to 1% strain) is far more limiting in design for high temperature integrity than tensile properties The creep–rupture
Table 3 Composition of various commercial or advanced/developmental austenitic stainless steel alloy grades and types (wt%)
(0.03C, <0.14N)
(0.03C, <0.14N)
CF3MN 17–21 9–13 2–3 <1.5 <1.5 <0.03 – – <0.1N (recommended)
1000
100
NRIM 316SS
9 heats
LMP LMP 10 000
10
18 000 20 000
Larson–Miller parameter
22 000 24 000 26 000
Figure 4 A plot of creep–rupture stress as a function of Larson–Miller parameter (LMP) for nine heats of SA 316 austenitic stainless steel tubing tested by the National Research Institute for Metals (now NIMS) in Japan LMP 10 000 represents data for rupture after 10 000 h LMP ¼ (T[ C]þ 273) (20 þ log t r ), where T is creep testing temperature and t r is the creep–rupture life in hours Reproduced from Data sheets on the elevated temperature properties of 18Cr–12Ni–Mo stainless steels for boiler and heat exchanger tubes (SUS 316 HTB), Creep Data Sheet No 6A; National Research Institute for Metals: Tokyo, Japan, 1978.
Trang 6strength of SA 316 in Figure 4 is comparable to
creep–rupture strength of 347 steel inFigure 5, and
both have less creep strength than 347HFG, a steel
containing more Nb and C (Table 3) Types 304 and
316L steels would have less creep strength than 316
steel By comparison, the triply stabilized (additions of
Ti, V, and Nb) HT-UPS steel has outstanding creep–
rupture resistance at 700–800C, comparable to that
of the solid-solution Ni-based alloy 617 A more direct
comparison of creep resistance at 700C and 170 MPa
is shown in Figure 6 For this creep–rupture
condi-tion, SA 316 ruptures after about 40 h, whereas the SA
HT-UPS steel resists creep and rupture until
18 745 h.4,7 For elevated temperature creep behavior
of heat-resistant stainless steels with additions of Ti
and Nb, processing conditions are also important,
including prior cold-strain and the SA temperature
The creep resistance of SA 304 and 316 steels is not
affected significantly by different annealing
tem-peratures, and both steels have less creep resistance
in the 10–30% CW condition By contrast, 347HFG
and HT-UPS steels benefit dramatically from higher
solution annealing temperatures (1050–1100C
com-pared to 1150–1200C) and small amounts of CW,
because these enhance the formation and stability of
nano-dispersions of MC carbide precipitates, which
are responsible for their high-temperature creep
resistance.4,7,10,11
300
Temperature for 100 000 h rupture life ( ⬚C)
HT-UPS
200
170
130
100
70 50
30
21 000 22 000
Commercial 347 [ref]
NF709
Alloy 617
Super304H TP347HFG
Trace from commercial austenitic stainless steels data
23 000
LMP {=(T [ ⬚C] + 273)(C + log trupture [h]), C = 20}
24 000 25 000 26 000 27 000 28 000
Figure 5 Creep–rupture resistance of high-temperature, ultrafine precipitate-strengthened steel compared to several commercial heat-resistant stainless steels and alloys.
700 ⬚C/170 MPa
100 000
10 000 1000
100 10
1
Austenitic stainless steel Figure 6 Direct comparison of creep-resistance of D9 and high-temperature, ultrafine precipitate-strengthened steels Adapted from Swindeman, R W.; Maziasz, P J.; Bolling, E.; King, J F Evaluation of Advanced Austenitic Alloys Relative to Alloy Design Criteria for Steam Service: Part 1 – Lean Stainless Steels; Oak Ridge National Laboratory Report (ORNL-6629/P1); Oak Ridge National Laboratory: Oak Ridge, TN, May 1990; Data sheets on the elevated temperature properties of 18Cr–12Ni–Mo stainless steels for boiler and heat exchanger tubes (SUS 316 HTB), Creep Data Sheet No 6A; National Research Institute for Metals: Tokyo, Japan, 1978; Teranishi, H.; et al In Second International Conference on Improved Coal Fired Power Plants; Electric Power Research Institute: Palo Alto, CA, 1989; EPRI Publication GS-6422 (paper 33-1); Swindeman, R W.; Maziasz, P J In Creep: Characterization, Damage and Life Assessment; Woodford, D A., Townley, C H A., Ohnami, M., Eds.; ASM International: Materials Park, OH, 1992; pp 33–42.
Trang 72.09.2.4 Precipitation Behavior During
Elevated Temperature Aging
Generally, austenitic stainless steels that have no
d-ferrite stay austenitic from room temperature up
to about 550C, at which temperature they can start
to experience the effects of thermal aging Aging
causes the alloy to decompose from a solid solution
into various carbide or intermetallic precipitate phases
and a more stable austenite phase The decomposition
of a quaternary Fe–Cr–Ni–Mo alloy, typical of type
316 stainless steel at 650C, is shown inFigure 7, and
the time–temperature–precipitation (TTP) diagrams for aging of SA behavior of type 316 and 316L stainless steel at 500–900C are shown inFigures 8 and 9.12,13 For typical light water reactor (LWR) or fusion reactor applications, such high temperature aging behavior is not too important, but it does become important for understanding irradiation-induced or -produced precipitation behavior for FBR irradiation of compo-nents at temperatures 400–750C As indicated in
above tend to produce precipitation of Cr-rich M23C6
in the matrix and along grain boundaries, while expo-sure at 600–750C eventually also produce precipita-tion of M6C, Laves (Fe2Mo), and s (FeCr) phases.12 Precipitation kinetics of these phases appears maxi-mum at 750–850C, and then at temperatures above 900–950C, none of these phases forms The lower
C content of 316L accelerates and shifts the formation
of intermetallic phases relative to 316 steel, as indi-cated in Figure 9 Additions of Ti or Nb cause the formation of MC carbides at the expense of the Cr-rich M23C6carbides, depending on whether the steel
is fully stabilized or not, but can also accelerate the formation of intermetallic phases, such as s or Laves
If d-ferrite is present in the alloy, it generally rapidly converts to s-phase during aging CW effects tend to accelerate the formation and refine the dispersion of carbides, but they can also significantly enhance the formation of intermetallic phases at lower tempera-tures, particularly in 20% CW 316.12–15 However, careful alloy design and compositional modification
316 composition range
Ni–50Fe 50
40
30
20 RIS at
sinks
RIS between sinks 10
0
Cr (wt%)
Ni (wt%)
Fe
(100)
a + g + s
a + s s
g + s g
a + g
a
Figure 7 Fe–Cr–Ni–X phase diagram at 650C X ¼ Mo.
Reproduced from Maziasz, P J.; McHargue, C J.
Int Mater Rev 1987, 32(4), 190–219.
1100
1000
900
M23C6+ c + s
M23C6+ Laves + c
M23C6+ Laves + c + s
M23C6+ Laves
M23C6+ M6C + Laves
M23C6+ M6C + Laves + s
M23C6+ M6C + α -ferrite
M23C6
M23C6+ c
800
700
600
500
400
Time (h)
s
s + c
Data, Weiss and Stickler Extrapolations by Maziasz Data, Maziasz Data, Stoter
Figure 8 Time–temperature–precipitation phase (TTP) diagram for SA 316 thermally aged Reproduced from Maziasz, P J.; McHargue, C J Int Mater Rev 1987, 32(4), 190–219.
Trang 8of certain austenitic stainless steels, such as the
HT-UPS steels, can result in alloys resistant to the
forma-tion of s-phase during aging or creep for up to
60 000 h or more The various precipitate phases that
form in 300 series austenitic stainless steels during
thermal aging or creep are listed below, with some
information on their nature and characteristics.12,14,15
M23C6– fcc, Cr-rich carbide, that can also enrich
Mo, W, and Mn, but is generally depleted in Fe, Si,
and Ni relative to the 316 alloy matrix
M6C – diamond-cubic phase that can be either
a carbide (M6C – filled, M12C – half-filled) or a
silicide phase (M5Si – unfilled), depending on how
carbon fills the atomic structure It is generally
enriched in Si, Mo, Cr, and Ni relative to the 316
alloy matrix
MC – fcc Ti- or Nb-rich carbide The Ti-rich MC
phase can also be very rich in Mo, or V and Nb, and
may contain some Cr, but tend to contain little or
no Fe, Si, and Ni The Nb-rich MC is a fairly pure
carbide phase that can enrich in Ti, but does not
usually contain any of the other alloying elements
in the 347 or 316 alloy matrix
Laves – hexagonal Fe2Mo-type intermetallic phase
Fe2Nb and Fe2W can also be found in steels
contain-ing those alloycontain-ing additions Phase tends to be highly
enriched in Si and can contain some Cr but is
generally low in Ni relative to the 316 alloy matrix
s – body-centered-tetragonal intermetallic phase,
consisting of mainly Cr and Fe It can be enriched
somewhat in Mo, but is depleted in Ni relative to the 316 alloy matrix
w – bcc intermetallic phase, enriched in Mo and
Cr, and containing mainly Fe, and depleted in
Ni relative to the 316 alloy matrix
FeTiP or Cr3P – hexagonal or tetragonal phos-phide compounds that can be found in stainless steels containing higher levels of P FeTiP is found
in the HT-UPS steels during aging
2.09.2.5 Corrosion and Oxidation Behavior With regard to general corrosion and oxidation, stainless steels with 16–18% Cr passivate and have good resistance to aqueous corrosion and various types of other acidic or corrosive environments at room temperature and up to about 200–300C.2 Additions of molybdenum give type 316 better resis-tance to pitting and acidic attack Effects of stress can aggravate corrosion resistance, and types 304 or
316 processed to have Cr-carbides precipitated along grain boundaries can suffer from stress-corrosion-cracking (SCC), which causes grain-boundary stress-corrosion-cracking
at reduced ductility to embrittle the steel Lower carbon steels (304LN, 316L) tend or reduce or elimi-nate SCC, as do the stabilized stainless steel grades such as 321 and 347, which form TiC or NbC carbides
to prevent Cr-carbide precipitation at grain bound-aries Exposure to supercritical water at 300C and above can be very corrosive, and cause oxidation of
1100
M23C6+ c
M23C6+ c + s
M23C6+ h
M23C6+ h + M6C
M23C6+ c + h
M23C6+ c + h + s
M23C6
1000
900
800
700
600
Time (h)
500
400
Figure 9 Time–temperature–precipitation diagram of solution-annealed 316L stainless steel during thermal aging Dashed lines represent a lower solution anneal temperature (1090C vs 1260C) Reproduced from Weiss, B.; Stickler, R Metall Trans 1972, 4, 851–866.
Trang 9austenitic stainless steels.16Generally, 300 series
aus-tenitic stainless steels have minimal oxidation in air at
500C and below, but oxidation and the protective
behavior of chrome-oxide scales become a concern at
550–600C and above Finally, 300 series steels such
as types 304 and 316 tend to show little or no corrosion
and behave quite well in liquid-metal sodium
envir-onments at 650C and below More detailed
informa-tion on austenitic stainless steels and their corrosion
behavior in aqueous environments, oxidation at
ele-vated temperatures, and behavior in liquid metals
such as sodium is available in other chapters of this
publication, or elsewhere
2.09.3 Summary of How Properties
Can Change During Irradiation
Other chapters in this volume present more details
on the fundamental nature and aspects of the primary
damage state in irradiated metals and alloys, and
on the detailed effects of irradiation on mechanical
properties’ behavior This chapter simply highlights
some changes in microstructure caused by fission or
fusion reactor neutron irradiation, and the changes in
properties that they cause in 300 series austenitic
stainless steels, to facilitate easy comparison to the
unirradiated behavior properties described above
Various other sections of this volume deal in far
more detail with the effects of irradiation in various
kinds of alloys
In LWRs at 20–250C, the interstitials migrate
freely to sinks, while the vacancies or their clusters
are relatively immobile, so this has been termed the
‘low-temperature regime’ of microstructural
evolu-tion in austenitic stainless steels.14In sodium-cooled
FBRs, temperatures are not lower than the sodium
coolant, so they are typically 300–350C or above,
which is termed the ‘intermediate-temperature
regime,’ and both vacancy and interstitial defects
can migrate to sinks Transmutation-produced helium
atoms are another form of primary radiation damage
that varies with reactor environment (high in LWR
and magnetic fusion reactor (MFR) systems, low in
FBR systems) Thermal neutrons produce helium in
austenitic stainless steels from boron atoms directly,
and by a two-step reaction with nickel atoms.17
He/dpa ratios for LWR systems can be very high,
over 100 appm He/dpa, while in mixed-spectrum
fission reactors used for radiation-effects studies on
materials, the ratios vary from 1 to 70 appm He/dpa
MFRs, with 14 MeV neutrons from D–T fusion
reactions, have linear He/dpa ratios of about 14 appm dpa1in a stainless steel first-wall component The FBR reactors with mainly fast fission neutron spectra produce very low He/dpa ratios of 0.1–0.5 appm He/dpa in austenitic stainless steels Irradiated mate-rials properties data discussed in the remainder of this section is mainly from LWR or mixed-spectrum fission reactor facilities used to study irradiation effects for MFR applications, so they have relatively high He/dpa ratios as well as a wide range of irradia-tion temperatures
The major effects of irradiation in mixed-spectrum fission reactors, such as Oak Ridge Research Reactor (ORR) or High Flux Isotope Reactor (HFIR),
on mechanical properties in the low-temperature regime are dramatic hardening (increased YS) and reduced ductility in SA and 316 and Ti-modified
316 stainless steels, and more modest hardening and ductility reduction in 20–25% CW steels The increased YS for irradiated SA steels are illustrated
250–300 MPa YS in the unirradiation condition, and 50% or more total elongation at room temperature and up to 250–300C, but irradiation increases the
YS to 600–800 MPa or more, and reduces ductility
to 10% or less However, the fracture mode in this irradiation temperature regime still remains ductile.8,9,19 After irradiation, 20–25% CW steels have YS of 800–1000 MPa, and less ductility, but still retain ductile fracture This is an important feature to note, and despite transmutation-produced helium levels of 1000–2000 appm, they do not embrittle, because helium and vacancy complexes are immobile
in this temperature regime However, most tensile test-ing results are in vacuum or air, and radiation-induced sensitization in water is not found after irradiation at 20–200C, but does become an embrittling factor to consider for irradiation above 300C.20
Irradiation-induced hardening of austenitic stain-less steels at room temperature to<250C is caused
by the microstructural changes produced by irradia-tion in this low-temperature regime Effects of alloy composition are small in this regime, but the effects
of processing condition prior to irradiation (SA or 20–25% CW) are very large Both SA and 25% CW steels, like 316 or Ti-modified 316, have very dense dispersions of ‘black-spot’ interstitial loops (2–4 nm diameter) uniformly within the grains,2,14,21 as illu-strated for 25% CW Ti-modified steel inFigure 11 However, the SA steels also have larger (10–50 nm) diameter Frank (faulted) interstitial loops and no network dislocations, whereas the 25% CW steels
Trang 10have a recovered dislocation network and virtually no
large Frank loops (Figure 12) These microstructural
effects directly reflect the fact that interstitial defects
are main point defects migrating freely to sinks in
this temperature regime Large Frank loops cannot
nucleate and grow until the concentration of network dislocations is below some critical concentration This also affects mechanical behavior, because the
‘black-dot’ and larger Frank loops are sessile until they unfault, whereas the network dislocations can
60 ⬚C
400 ⬚C
20 nm Figure 11 Transmission electron microscopy of
black-dot loops in 25% CW PCA irradiated in ORR at 60
and 400C Reproduced from Maziasz, P J J Nucl Mater.
1992, 191–194, 701–705.
0
200 400 600 800 1000
Temperature ( ⬚C)
3–20 dpa
316 and PCA steels
Unirradiated
Figure 10 Yield strength as a function of irradiation temperature for SA 316 and PCA in various reactors Reproduced from Pawel, J E.; Rowcliffe, A F.; Lucas, G E.; Zinkle, S J J Nucl Mater 1996, 239, 126–131.
10 16
1015
10 14
Irradiation temperature ( ⬚C) 200
Network
Larger frank loops
-2)
‘Black-dot’
loops
Total
As-cold-worked
25% CW PCA ORR (6 J/7 J) 7.4 dpa
Figure 12 Plot of dislocation density versus irradiation temperature for various components of dislocation structure for 25% CW PCA irradiated in ORR at 60–400C Reproduced from Zinkle, S J.; Maziasz, P J.; Stoller, R E.
J Nucl Mater 1993, 206, 266–286.