27 - 725Materials for Nuclear EnergyIrradiation Embrittlement of

Pressure Vessel Steels (Ferritic)

Tony (A.D.) Rollett, Carnegie Mellon University

Spring 2019

Updated 29th Jan. 2019

Most slides taken from a lecture by Berquist & Burke on “Pressure Vessel Steels”

Please acknowledge Carnegie Mellon if you make public use of these slides

Outline1. Radiation Damage Event2. Atom Displacement3. Damage Cascade4. Point Defect Formation & Diffusion5. Loss of, Sinks for Point Defects6. Radiation Induced Segregation7. Radiation Induced Dislocations8. Radiation Induced Voids, Bubbles9. Radiation Effects on Phase Stability10. Ion Bombardment, Implantation11. Surrogates for Neutron Irradiation12. Radiation Effects on Hardening, Deformation13. Radiation Effects on Fracture Toughness, Embrittlement14. Radiation Effects on Creep and Growth15. Radiation Effects on Corrosion: emphasis on Stress Corrosion

Cracking (SCC)

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Charpy Test• The Charpy test uses a square bar

with a small notch in it.• The further the pendulum swings after

breaking the specimen, the less energy was absorbed in the impact, and vice versa.

• Higher toughness results in higher energy absorbed.

• The test is effectively a dynamic test because the strain rates are much higher than in a fracture toughness test.

• Brittle materials exhibit low toughness;conversely, ductile materials exhibit high toughness.

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Fractography• These micrographs contrast the appearance of ductile

and brittle fractures at the microstructural scale.

Dowling

Brittle(cleavage)

Ductile(tearing)

Toughness vs. Temperature• Most metals are ductile at high temperature

and brittle at low temperature. The transition in toughness is, as we shall see, quite sharp so we can define a Ductile-to-Brittle Transition Temperature, DBTT.

• FCC metals are the exception because they do not exhibit a DBTT but (typically) maintain high toughness down to cryogenic temperatures.

• BCC metals all exhibit a DBTT, which for Fe is around room temperature. This means that one must be careful with alloying to avoid having a too high DBTT.

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This WW-2 era “Liberty” ship, with a welded steel hull, failed by brittle propagation of a crack while moored in dock! The fracture toughness of the steel was too low for the design.

General Points

• The most serious issue of all is the increase in the Ductile-to-Brittle Transition Temperature or DBTT in (ferritic) pressure vessel steels, which is determined at 41 J (slightly higher in Russia). Note the shift to the right (higher T) and decrease in upper shelf energy in the graph.

• This is a serious issue because, to first order, the increase in temperature is linear in the dose (dpa). Operating a pressure vessel at a temperature where the fracture toughness is low implies that a catastrophic burst is at least potentially possible.

• This issue was also addressed in the NATO conference article by Odette.• As will be addressed by Fyfitch, stainless steels and nickel alloys also experience loss of

ductility, especially once swelling has started.

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• Along with hardening, irradiation of metals results in losses of ductility and decreases in fracture toughness.

JOM • July 200918 www.tms.org/jom.html

the status of the recent $T models, with emphasis on the major outstanding is-sues. In summary: • $T depends on combined effects

of neutron flux (F), flux spectrum, fluence (Ft), irradiation tempera-ture (Ti), alloy composition (Cu, Ni, Mn, P), and start-of-life mi-crostructure, or product form.1,2,7 Due to scatter and clumped and confounded distributions of vari-ables, however, the PREDB lacks the resolution to statistically iden-tify a “best” $T model. Properly supplementing the PREDB with high-resolution IVAR and other qualified test reactor data will be needed to significantly improve the reliability of future $T mod-els.

• Existing $T models cannot reliably extrapolate to high fluence levels of up to ≈1020 n/cm2 pertinent to pressurized water reactor extended life conditions, since 99% of the PREDB data is at <5×1019 n/cm2. Thus, high fluence data from other surveillance programs and test re-actor irradiations must be used to develop improved $T models for extended life. Figure 1b shows predicted minus measured re-siduals for a successful low flux PREDB based $T model1 applied to a large independent body of higher flux test reactor data, com-pliled by M. EricksonKirk.8 (M. EricksonKirk provided test reac-tor database to the authors in the form of an Excel spreadsheet. Er-icksonKirk also derived a number of $T models based on systematic fits to the PREDB and selected subsets of the test reactor data-base, as described in detail in the draft NRC report Technical Basis for Revision of Regulatory Guide 1.99: NRC Guidance on Methods to Estimate the Effects of Radia-tion Embrittlement on the Charpy V-Notch Impact Toughness of Re-actor Vessel Materials. A sum-mary of this work is presented in Reference 8.) The large negative residuals that increase with fluence show that the model systematical-ly and significantly underpredicts $T. Thus, reliably modeling high fluence embrittlement at the much

PREDB-based $T models. Since RG 1.99-2 neither provides a good fit to the existing PREDB, nor reflects current understanding of em-brittlement mechanisms, research has

continued to develop improved $T models, including one that has already been used in regulatory assessments of pressurized thermal shock.1 In the fol-lowing sections we will briefly review

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Test Temperature (ºC)

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ture

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HSSI Weld 72 W1st and 2nd Batches

1 TC (T) 1st Batch1 TC (T) 2nd Batch1 TC (T) (To = 54.0ºC)MPC PCVN (To = 75.0ºC)

Master Curves

Ft (1019 n/cm2)

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Square Symbols - MPA DataCircle Symbols - ORNL Data

KS-01 WeldIrradiated at 288ºC0.8 s 1019 n/cm2 (E>1MeV)

c

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Figure 1. (a) Charpy impact energy-temperature plots for the unirradiated and irradiated KS-01 weld (0.37 wt.% Cu, 1.23 wt.% Ni).3 (b) EONY model predicted minus measured $T residuals for the test reactor database assembled by M. EricksonKirk.8 (c) Fracture toughness vs. test temperature for HSSI Weld 72W, comparing T0 with more than 200 small precracked Charpy specimens (data not shown) tested as part of a multi-laboratory Materials Properties Council study with 36 tests of larger 1TC(T) specimens; the measured T0 is higher for the larger specimens.17

Odette & Nansted JOM (2009) 61 17-23

Upper shelf energy

41 J

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Radiation Effects

• Focus on metals (steel)• Effects include:

– Hardening• Increase in yield stress & UTS

– Embrittlement & Fracture• Reduced plastic or creep deformation/reduced ductility• Brittle fracture with rapid crack growth

– Void Swelling• Interstitial loops cause expansion; if vacancies agglomerate into voids,

then no contraction corollary• Requires sufficient thermal activation (energy) for agglomeration; not so

important for recombination– Irradiation Creep

• Can be “irradiation enhanced” or “irradiation induced”

• Effects depend on level of irradiation & material temperature

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Summary of Effects

Integrated fast neutron* flux refers to fission reactor neutron spectrum but can be related to high energy fusion neutrons too

Important for fission fuel & cladding, and for fusion first wall and blanket

*See Supplemental slides

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Summary of Effects

Thermal & epithermal neutrons à more prevalent in reactor structure

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Radiation Hardening

Pressurized Thermal Shock (PTS)• During normal operation, embrittlement of the vessel wall occurs by fast neutron

damage, which results in reduction of KIc, or alternatively, an increase in the ductile-to-brittle transition temperature (∆T) as measured in Charpy tests.

• Brittle fracture is not a problem during normal operation because the temperature is high (~300�C) and the main stress on the vessel is due only to residual stresses and the membrane (hoop) stress due to the internal pressure in the vessel.

• If cold water from the emergency core cooling system (ECCS) should be injected into the system (because of a LOCA= Loss of Coolant Accident) near the end of the vessel lifetime when KIc is lowest due to irradiation embrittlement, brittle failure is possible.

• The rapid reduction in the inner wall temperature during ECCS creates tensile thermal stresses which add to existing tensile stresses (from the pressurization).

• Small flaws (cracks) on the inner wall of the vessel may have grown throughout the reactor lifetime due to fatigue during shutdown and startups.

• If the increased crack size and the corresponding increase in the applied stress intensity factor (KI) exceeds KIc, the crack will grow rapidly.

• The crack may not penetrate the entire vessel wall because temperature, stress and irradiation dose change through the wall thickness. The crack will stop when the applied KI reaches the crack arrest toughness KIa. [Wirth]

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Remedies for PTS Problem• [from Wirth notes]• For new pressure vessels, minimize the Cu content of

the steel in particular, also avoid high Ni and Mncontents.

• Russian reactor vessel codes were more concerned with Cu and P.

• For old, existing pressure vessels, take following actions:- anneal critical components- manage the position of high burn-up versus new fuel elements- use reflector parts that can handle neutron irradiation such as tungsten- heat make up water (not practicable?!)

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Microstructural evolution: ferritic steels

• Here we discuss the response of ferritic steels, such as those used for pressure vessels

• For T<~350 oC– Depleted zones & vacancy clusters are stable “black dots”,

which are clusters of solute atoms (esp. Cu, Mn, Si, Ni) too small to resolve in a transmission electron microscope

– Increased neutron fluence à increased “black dot” density

• For T>~350 oC– Microstructure is completely different!– Point defects mobile– Dislocation loops & voids form

Materials Properties of Ferritic SteelsKey Embrittlement Property : Toughness

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Ductile-to-Brittle Transition Temperature or DBTT is determined at 41 J

Mechanical Properties of Pressure Vessel Steels

• Impact properties of a standard sized Charpy Test • Key parameter is Energy Absorbed during fracture• Perform tests over a range of temperatures –plot provides a standard curve

ØSigmoidal shapeØUpper Shelf EnergyØLower Shelf EnergyØ30 ft-lb “Ductile to Brittle Transition Temperature”

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Ferritic RPV Steels:Effects of Irradiation/Service Degradation

• Long term degradation of RPV Steels has been observed – shift in ductile to brittle transition and reduction of upper shelf energy

• Degradation occurs by “hardening and embrittlement”• Hardening is due to

– irradiation induced crystal damage/defects (e.g. vacancies, interstitials, dislocation loops and debris, )

– formation of nanoscale clusters enriched in solute– Current thinking points to Mn-Si-Ni precipitates that only form

at low temperatures with kinetics enhanced by irradiation

• Hardening of the material matrix will enhance any existing susceptibility to embrittlement.

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Irradiation Hardening of Low Alloy SteelsMicrostructural Features

• All features are on sufficiently fine scale that….• High resolution microscopy and microanalysis are required to observe

defects• Multiple techniques are required to separate out size and chemistry effects• Heavily dislocated structures obscure observation of other features• Field Ion Microscopy (FIM) results have identified “solute enriched clusters”

rather that “copper precipitates” as solute hardening• Voids and dislocation structures observed by Transmission Electron

Microscopy (TEM)• Advanced Analytical Electron Microscopy (AEM) can now be used to

resolve local chemistry variations• The Atom Probe*, combined with atomistic modeling, has been particularly

useful for understanding the generation of solute clusters in steels.

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* 3DAP, successor to the field ion microscope, where atoms are removed from the tip of a sharp needle-shaped specimen one at a time, sometimes with the aid of laser light, and analyzed in a mass spectrometer

Microstructures of Irradiated Pressure Vessel Steels – Solute Clustering

T. Shikama on KINKEN Research Highlights 2007 Web site

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Ferritic RPV Steels: Irradiation/Service Degradation

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solute-relatedhardening

Effect of Irradiation and Annealing on Hardness of A533 Gr 4 (3.5Ni) RPV Steel(M.G. Burke et al., in “Effects of Radiation on Materials” ASTM STP-1447, 2002))

Solute cluster related hardening cannot be annealed out, unlike vacancy related hardening

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Solute clusters• Why is Cu a problem? Odette [1984 Scripta] points out that the

solubility of Cu at 290�C is about 0.0002%, which is far less than the typical level of 0.01% encountered. The ratio is 50 which implies a large driving force for precipitation. He further notes that the cleavage stress (seen at low temperatures in the lower shelf) is unlikely to be sensitive to irradiation because it is controlled by carbides that are too coarse to be affected by irradiation. Intergranular embrittlement is also absent in this class of steels. Therefore it is increases in strength that lead to increases in the DBTT.

• Modern analysis clearly points to fine precipitates of Mn-Ni-Si as being the culprits for hardening [2014 Acta Wells et al.].

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Atom probe reconstructions

– Both N and hri increase with Cu content (Fig. 3b). Nalso increases with Ni content, except in the Cu-freesteels. It should be noted that at these small f and hri,the uncertainties in N are larger.

– The MNS cluster compositions are generally similar inthe medium- and high-Ni content steels (LG,LH,LI,LC, LD) with fractional averages and standarddeviations of: 0.31 ± 0.04 Mn, 0.47 ± 0.01 Ni and

0.22 ± 0.03 Si. The precipitate Ni fraction is higherand the Si is lower in the highest-Ni content, Cu-freesteel (CM6) averaging 0.26 Mn, 0.59 Ni and 0.15 Si.

The most important observation for the high-fluencecondition is that Cu and Ni play a combined role in medi-ating f, although Cu seems to have a stronger influenceover the range of Ni compositions studied here. While

Fig. 2. Atom maps for the highest-Ni content, Cu-free (top), and high-Ni–Cu content (bottom) alloys irradiated to high fluence.

Table 5Precipitate compositions and hri, N and f at high fluence (G1).

Alloy Precipitate relative compositions (at.%) Nominal Fe** hri (nm), N (m!3), f (%)

Cu ± Ni ± Mn ± Si ± hri ± N ± f ±

LC* 25.7 1.4 35.1 5.0 23.6 7.2 15.7 3.8 60.8 1.15 0.10 9.2E+23 1.4E+23 0.58 0.08LD 22.2 0.7 37.8 0.5 24.6 1.2 15.5 0.7 58.6 1.13 0.02 11.5E+23 0.7E+23 0.68 0.04LG 0.5 0.1 46.8 2.3 25.5 3.6 27.2 1.9 53.3 0.72 0.04 5.3E+23 1.5E+23 0.08 0.02LH 14.8 0.9 39.8 2.5 25.1 1.8 20.3 0.1 58.7 0.92 0.05 4.9E+23 0.5E+23 0.16 0.01LI 24.2 1.0 34.2 0.5 27.4 0.8 14.2 0.4 57.2 1.10 0.03 6.9E+23 0.6E+23 0.37 0.04CM6* 0.1 0.1 58.9 5.0 25.6 7.2 15.5 3.8 49.6 0.70 0.10 4.6E+23 3.0E+23 0.07 0.04

* Note that measurements for LC and CM6 from this irradiation condition came from a single APT run of LC and CM6. The uncertainty given for thesealloys is twice the largest uncertainty of the other alloys. The CM6 mole fraction uncertainty was twice the uncertainty of the LG volume fraction as theyhad similar measured f.** Assumed to be an artifact, but provided for alternative interpretations.

P.B. Wells et al. / Acta Materialia 80 (2014) 205–219 211

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2014 Acta mater.Wells et al.

Toughness : Impact Testing

• Impact properties of a standard sized Charpy Test • Key parameter is Energy Absorbed during fracture• Perform tests over a range of temperatures – plot provides a standard curve

ØSigmoidal shapeØUpper Shelf EnergyØLower Shelf EnergyØ30 ft-lb “Ductile to Brittle Transition Temperature” (DBTT)

• Differences in MaterialsØChange in Upper Shelf Energy (USE)ØShift in DBTT

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Change in Toughness vs. Hardening• The upper figure is a

Davidenkov diagram, showing the effect of hardening on embrittlement.

• The second fig. shows the remarkably (near-)linear relationship between the fractional decrease in the upper shelf energy, plotted versus the hardening increment.

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Effect of Cu

• An A-302B steel was made with various levels of Cu and Ni. See the ref. for full details.

• Note that the low copper steel, upper panel, shows a much smaller increase in DBTT compared to the high Cu+Ni steel, lower panel.

• Recall that European PWRs+BWRs generally have lower Cu contents in their PV steels.

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Recovery treatments

• In the same paper, the effect of post-irradiation heat treatments was investigated.

• In this case, the DBTT was significantly reduced, although not completely back to the un-irradiated state.

• The latter point is the same as that noted on slide no. 16

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Summary• Irradiation brings about increases in hardness and yield stress in

structural materials used in reactors.

• Reactor Pressure Vessel (RPV) Steels are ferritic and exposed to

moderate irradiation at mildly elevated temperatures – the concern

is the steady increase in hardness and consequent decrease in

Upper Shelf Energy (USE) and increase in the Ductile-to-Brittle-

Transition Temperature (DBTT). The key scenario is a Loss-of-

Coolant accident (LOCA) coupled with Pressurized Thermal

Shock, leading to catastrophic crack propagation.

• The magnitude of hardening depends on the composition: Cu in

steel (a tramp element, generally) strongly increases the radiation

hardening.

• The main contributors to hardening are voids, dislocation loops,

bubbles and network dislocations. However, clustering of solutes

also appears to contribute to hardening (but is hard to quantify).

• The hardening dependence on dose rate is often sub-linear.

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Supplemental Slides

http://en.wikipedia.org/wiki/Neutron_temperature

• Neutron energy distribution ranges

• Moderated and other, non-thermal neutron energy distributions or ranges are listed in the table below:

• Fast neutrons have an energy greater than 1 eV, 0.1 MeV or approximately 1 MeV, depending on the definition.

• Slow neutrons have an energy less than or equal 0.4 eV.• Epithermal neutrons have an energy from 1 eV to 10 keV.• Hot neutrons have an energy of about 0.2 eV.• Thermal neutrons have an energy of about 0.025 eV.• Cold neutrons have an energy from 5x10−5 eV to 0.025 eV.• Very cold neutrons have an energy from 3x10−7 eV to 5x10−5 eV.• Ultra cold neutrons have an energy less than 3x10−7 eV.• Continuum region neutrons have an energy from 0.01 MeV to 25 MeV.• Resonance region neutrons have an energy from 1 eV to 0.01 MeV.• Low energy region neutrons have an energy less than 1 eV.

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Effects on Mechanical Properties• Defects produced by neutron irradiation (coupled

with time and temperature in some cases):1. Point defects

2. Impurity atoms

3. Vacancy clusters (depleted zones)

4. Dislocation loops

5. Dislocation lines

6. Cavities (voids & He bubbles)

7. Precipitates (e.g. carbides)

• 1,2 do not (directly) cause hardening; 3-7 do

• Two strengthening mechanisms:– Source hardening (increased stress required to initiate

dislocation motion)

– Friction hardening (moving dislocations impeded by obstacles that are close to or in the slip plane)