course material

1. CHEM1040 General Chemistry I
Answers:
Introduction
One of the critical problems that affect the cracking of light water reactors core components
is the irradiation-assisted stress corrosion cracking (IASCC) 7. Austenitic stainless steel was
irradiated in reactors so that studies on IASCC could be conducted. Irradiated
microstructure and the role of microchemistry in IASCC were investigated in this study 3.
The influence of neutron irradiation on stainless steel components inside a reactor vessel
leads to changes during plant operations. The radiation primarily induces new point
defects 1. The defects migrate, interact with the original microstructure in relation to the
radiation temperature, and stable complexes of the defects are formed 1. The stainless steel
undergoes processes like swelling, toughness, losses ductility, and hardens, which are
defects of microstructure, as well as undergoes radiation-induced segregation 1.
The studies conducted earlier used spectroscopic techniques for microchemical analysis for
characterization of grain boundary11. Oxygen and fluorine contamination led to
intergranular SCC 11. The failure of type 304L and 304 stainless steel core was explained by
the oxyflourine-assisted stress corrosion cracking11. Chromium, fluorine, and oxygen
strongly influenced intergranular stress corrosion cracking 11. A recent study investigated
the effects of large amounts of helium in a reactor on IASCC susceptibility 10. The results
indicated that with no significant displacement damage, helium does not lead to IASCC and
intergranular cracking in hydrogenated high-temperature water and high-temperature air,
respectively 10. The susceptibility for IASCC and irradiated microstructure was determined
recently for neutron-irradiated cold–worked 316 stainless steel with various damage
levels 12. IASCC threshold stress was 45% of the irradiated yield strength for neutron-
irradiated steel between 46.9 dpa and 125.4 dpa 12.
The various studies conducted have illustrated that increase in dose affects the degree of
intergranular stress corrosion cracking. Hardening, segregation, voids, and dislocation
loops are various ways through which the microstructure of alloys keeps changing as they
all correlate with increased susceptibility to cracking 3. In light water reactors (LWR), the
irradiated microstructure is usually dominated by the evolution and formation of faulted
dislocation loops 7. This leads to increased yield strength and hardness and a reduction in

2. uniform elongation.
The changes that are related to yield strength may cause effects in the cracking of austenitic
alloys. The crack growth rate as well increases with an increase in yield strength in stainless
steel in 288 ? BMR water 3. The occurrence of IASCC is contributed greatly by radiation
induces segregation. The irradiated microstructure may be affected by stacking fault energy
which varies depending on the composition of alloys. The determination of IASCC is
challenged by the fact that localized deformation, hardening, loop microstructure, and RIS
tend to saturate at similar doses, and they develop at almost the same rates7. At relatively
low doses, the IASCC susceptibility is dominated by mechanisms like hardening, relaxation,
RIS, and radiolysis 10. The amount of cracking and IASCC increases with exposure to
neutron 1. This is a study that is based on the effects of microchemistry and irradiated
microstructure on IASCC.
Summary
Irradiated-assisted stress cracking (IASCC) is the cracking propensity of an alloy in the
presence of irradiation in water that has a high temperature3. Reduced time for initiation of
crack or increase in growth of a crack compared to its unirradiated counterpart indicate the
accelerated cracking3. The water chemistry and material are both affected by radiation3.
Physical changes to microstructure like hardening and segregation that are radiation-
induced are the effects of Irradiation in austenitic alloy 1. Irradiation hardening and
radiation-induced segregation cause IASCC 4. The authors expound on the factors that cause
IASCC as outlined in the segments below.
Radiation-Induced Segregation
The non-equilibrium process occurring at point defect sinks during alloy irradiation at
intermediate temperature is known as radiation-induced segregation (RIS) 2. The unequal
participation of solutes in vacancy and interstitial fluxes to sink results from segregation
that is induced by solutes3. Depleting the element at the defect sink or enrichment of an
alloy element occurs when there is unequal participation of solutes. The association of the
interstitial flux with the solute also causes depletion and enrichment of elements at the
defect sink 3. Oversized elements deplete, whereas enrichment occurs to undersized
elements 3. The segregation in Fe-Cr-Ni alloys can be explained by the inverse Kendal
mechanism or vacancy exchange 3. A net solution flux away from or toward the boundary
results from an alloying element with vacancy flux 2. The exchange of a solute with vacancy
flux is one of the significant contributors of RIS in Fe-Cr-Ni alloys 2. The interaction of
undersized atoms with interstitial flux is another contributor of RIS in austenitic alloys like
Fe-Cr-Ni 2. The chromium is depleted at the grain boundaries while nickel is enriched,
whereas iron can deplete or enrich as it's depended on the composition of bulk alloy 2. The
experimental results below show the RIS of silicon, nickel, chromium, and phosphorous at
the grain boundary. The stainless steel was irradiated in a water reactor to several dpa at

3. 300? 2.
Figure 1: showing RIS of chromium, silicon, phosphorous, and nickel at the grain boundary
of irradiated stainless steel in a water reactor 2.
Irradiated Microstructure
At LWR service temperatures, the microstructure of austenitic alloys like stainless steel
changes rapidly. At very low doses, the point defect clusters are formed 3. Network
dislocation densities and dislocation loops evolve over several dpa with dose 3. This was
confirmed by an experiment conducted on 7 alloys that were irradiated to one and five dpa
at 288 ? 4. The voids and dislocation loops varied over the irradiation doses and alloys. A
possibility for formation of voids, He filled bubbles and precipitates in core components that
have been exposed to higher temperatures and dose 3. During radiation, the collapse of the
damage cascade may lead to the formation of defect clusters that may be of interstitial or
vacancy type 3. The high mobility of interstitials leads to the growth and nucleation of large
dislocated loops 3. The number density and loop population will grow until the population
saturates when interstitials and absorption of vacancies equalize. At 1 dpa, the saturation of
loop number density occurs, whereas the loop size evolves to 3-5dpa 3.
The loop size and number density depend on alloying elements and conditions of
irradiation. The number densities are in the order of 1×1023 m-3, and the loop size hardly
exceeds 20nm 3. This was contradicted by an experiment conducted to determine the effect
of alloy chemistry on loop size and density. The findings indicated no observed correlations
between alloy chemistry and loop size and density 5. However, the varying experimental
conditions may have suppressed the correlations 5. The reduction of loops as a defect sink
helps bubbles and voids grow3. The dislocation structure becomes the dominant
microstructure as the loop size and number density increase with temperature 3. Larger
Frank loops unfault is made from the dislocation structure that evolves into a network
structure 3. Frank loops are primarily a contribution to radiation hardening compared to
network dislocations and black spots at relevant temperatures of LWR 5.
The growth of second phases can be retarded or accelerated as well as new phases can be
produced, and existing phases can be modified by irradiation3. The radiation-induced
segments of minor and major elements can induce phase formations exceeding the limit of
solubility. However, the effect of stability and phase formation by radiation occurs above
400 ?. Irradiation at higher temperatures can lead to induction of lava phase, γ-Ni3Si, and G
phase formation. This was confirmed by an experiment that predicted second phase
molecules at 400 ? 5. The results obtained indicated that precipitation of G phases and γ'
was enhanced by dislocation density after irradiation 5.

4. Figure 2: showing a simulated evolution of precipitate in 316SS as a function of aging time
at 400? 5.
Radiation Hardening
Austenitic alloys like stainless steel are essential in structural materials. Irradiation
hardening is a great concern for reactor terminals used for long-term services and is placed
in environments exposed to radiation 6. Radiation hardening results from dislocation loop
microstructure. The dislocation network and loop react elastically upon the application of
stress, thus increasing the yield strength of the austenitic alloy. Indication hardness
measurement or tensile tests are used to determine the increase in yield strength 3. In cases
of proton radiation, implantation of almost 3×105 produced gas bubbles and vacancy
complexes that gave the dislocation motion a stronger barrier causing hardening 6. The
hardness obtained from Xe and H are studied, and protons produce a steep peak
displacement damage that is non-uniform. The saturation fluence of a proton is higher 6.
Figure 3: showing different hardness saturation obtained for Xe and proton radiation 6.
Deformation mode changes dramatically with increasing dose in addition to hardening.
Narrow channels which have been cleared of defects by preceding dislocations are localized
plasticity 3. The sharp reduction in uniform elongation and localized necking are caused by
intense shear bands that result from dislocation channeling3.
Characteristics Of IASCC
The IASCC of irradiated alloys is dependent on similar parameters as the IGSCC of
unirradiated alloys. Changes to the alloy and changes in water chemistry are factors
through which radiation affects IG cracking3.
Effects Of Radiation On Water Chemistry
The origin of radiation effects on water chemistry is represented by the formation of
reductants and oxidants3. The thermodynamics and electrochemical reactions are
controlled by corrosion potential, which is influenced by reductants and oxidants that are
formed as a result of radiation3. A sigmoidal decrease in corrosion potential is observed
when mass transport is limited by the reduction in oxidant levels. Materials pre-irradiated
by slow strain rate were studied. A decrease in average crack growth rate was observed
with decreasing potential of corrosion 3. Corrosion potential is the dominating factor in
unirradiated alloys, and IGSCC of irradiated that is controlled by ionic species and oxygen

5. content in water 3. This is confirmed by Anna by highlighting that water dissociates into
different radical products of reaction, ionic and molecular products under the influence of
radiation 1. The radical reaction products react to form O2, H2, and hydrogen peroxide
(H2O2) 1. The products of radiolysis are scavenged by the addition of hydrogen to reactor
water hence reducing the effect of radiolysis 1. In BMR, the products of dissociation of water
lead to an increased potential of corrosion 1.
Irradiated Microstructure And IASCC
The mechanisms of IASCC are categorized into effects of material and water chemistry. The
principal effects of material and water chemistry indicate a steep increase of kinetics of
environmental cracking with an increase in corrosion potential. There is a sharp increase in
the growth rate of crack for a solution with conductivities that are above 0.1 µS/cm in
unirradiated or irradiated cases. Solution conductivity and corrosion potential in constant
water chemistry conditions control the cracking rate3. IASCC is strongly influenced by the
cumulative radiation effect, which is illustrated by the increase in cracking susceptibility
with fluence in rare tests and slow strain3.
Radiation-induced segregation at grain boundaries mechanism of IASCC
IGSCC susceptibility in alloys like stainless steel and grain boundary chromium level show a
correlation where thermal sensitization causes grain boundary depletion. In austenitic
alloys, chromium depletion is an agent in IGSCC in oxidizing conditions. The boundary
becomes a location for oxide destabilization, and passivation by the formation of a
protective chromium oxide over the grain boundary is not possible when the concentration
of chromium grain boundary drops below the threshold level. The chromium depletion
profiles that result from precipitation reactions or RIS show a difference by having a smaller
width of RIS profiles by two orders of magnitude. Chromium depletion at the grain
boundary is insufficient but an important condition for the occurrence of IASCC 3. The
depletion of chromium in oxygenated water was seen to be sufficient to cause IASCC 9. This
is because post-irradiation annealing showed a change in IG cracking susceptibility in BMR
water while RIS was virtually unchanged 9. RIS is found to be dependent on GB structure,
the dose of irradiation, and the composition of alloy 5. The structure of GB has a strong
ability to suppress segregation to GBs under irradiation 5.
Dislocation Loop Microstructure And Radiation Hardening Mechanism Of IASCC
The dislocation of loop microstructure on alloy behavior is hardening which follows a
dependence on the square root of the product of loop size and loop number density
according to the hardening model of the dispersed barrier. Hardening is an essential factor
in the susceptibility of IASCC 3. Increased cold work leads to an increase of crack growth in
austenitic stainless steel that is not sensitized. Increased yield strength from cold work can
predict the susceptibility of cracking 3. Cracking susceptibility is promoted by cold-working

6. and subsequent deformation during SCC tests 8.
The parallel between IASCC irradiation and cold work effects indicates that hardening is an
essential factor in IASCC susceptibility. Cracking and hardening are reduced by increased
annealing, but the behavior of hardening cannot satisfactorily explain the complete and
rapid reversal of susceptibility in IASCC with annealing condition 3. The susceptibility of
cracking changes drastically before the dropping of hardness from its irradiated value
begins. The various set of experiments that were conducted by Hash revealed that
hardening was not the only factor in IASCC. This is because IASCC susceptibility was not
constant when hardness was the only sole factor. Cracking was also higher in lower doses
than in companion samples that had similar doses with no cold work. This indicates that
IASCC susceptibility can also be enhanced by cold work 3.
Deformation Based IASCC Mechanism
The IASCC proton irradiated samples over a wide range of doses have shown that nickel
alloys have high IASCC resistance. The hardening and dislocation microstructure is similar
in 18 chromium- 8 nickel alloys, and through a dose of 5 dpa, no voids were observed 3. The
higher content of nickel in stainless steel increases significantly the stacking fault energy
(SFE). A change in slip nature from a planar slip to a wavy slip results from increasing SFE 3.
The SCC resistance in stainless steel has been linked to SFE. In stainless steel, decreased SCC
susceptibility and increased area reduction correlated well with increasing SFE 3. The
rupture of an oxide could be caused by stress at grain boundary at lower fluences before the
occurrence of matrix material cracking leading to subsequent corrosion, exposure of the
material to the solution, and IASCC 3.
IASCC can also occur by deformation of the boundary leading to oxide film rapture causing
IASCC 3. Localized deformation is needed to break oxide film that covers grain boundary 7.
The array of dislocations irradiation causes the formation of dislocation channels. The
formed dislocation channels terminate or transmit at the grain boundary. A greater impact
on dislocation channels results when larger dislocations are involved since the number of
glide dislocations in the channel is proportional to the height of the channel 7.
The experiments conducted indicated that cold work was less effective than radiation
hardening in the induction of IGSCC 3. Slip bands and dislocation channels can influence
cracking. The grain boundary stacking fault energy (GB SFE) slightly increases with doses
for all alloys as there was no clear difference in dose dependence for any alloy 3. The GB SFE
is increased by nickel enrichment from RIS, whereas molybdenum and chromium depletion
lowers the value 3. The experimental results obtained indicated that lower SFE is found in
susceptible 304 alloys than 316 alloys that are resistant in the unirradiated condition 3.
This suggests that SFE at the grain boundary could have an influence on the determination
of IASCC susceptibility 3.

7. The Achievements Of The Study
The authors were able to identify that the inverse Kendal mechanism or vacancy exchange
could be used to explain radiation-induced segregation in Fe-Cr-Ni alloys. This was
confirmed by research that also identified that solute exchange with vacancy flux was a
significant contributor of RIS in Fe-Cr-Ni alloys, as well as the interaction of interstitial flux
with undersized atoms. The study also highlighted that point defect clusters are formed at
very low doses because, at LWR service temperatures, the microstructure of austenitic
alloys changes rapidly. This was verified by a study conducted on 7 alloys that were
irradiated to 1 and 5 dpa at 288 ?. The voids and dislocation loops varied over irradiated
doses and alloys. The study also highlighted that increment in yield test can be used to
indicate hardness measurements. This as well is authenticated by radiation hardening
results that were obtained from the dislocation of loop microstructure. The dislocation loop
and network reacted elastically upon the application of stress; hence the yield strength of
the austenitic alloy increased.
Ionic species and oxygen content in water control IGSCC of irradiated alloys observation
was confirmed by Anna, who highlighted that water forms radical products upon irradiation
that react to form O2, H2, and hydrogen peroxide (H2O2), increasing the potential of
corrosion. Chromium depletion was identified as an agent in IGSCCin oxidizing conditions.
Research by Onchi et al confirmed that chromium depletion in oxygenated water was
sufficient enough to cause IASCC. The various studies conducted have shown that factors
that lead to susceptibility of IASCC are dependent on each other, as the authors had
indicated.
Limitations Of Approach
This study is limited because localized deformation, hardening, loop microstructure, and
RIS tend to saturate at similar doses and develop at almost the same rates7. Therefore, the
occurrence of IASCC is dependent on factors that correlate with each other. The primary
factor causing susceptibility of IASCC was not identified. The SFE, deformation of the
boundary leading to the rapture of oxide film, cold work, and hardening all contribute to the
susceptibility of IASCC.
Conclusion
The change in hardness of alloys like stainless steel, microchemistry, and microstructure
are resultant from irradiation of stainless steel. The alloy composition and microstructure
that has been irradiated can affect the mode of deformation compared to a sample that has
not been irradiated. Deformation of the material at the grain boundary through the oxide
film rapture is a potential mechanism for irradiation-assisted stress corrosion cracking.
Boundary dislocations result in sliding of the grain boundary or localized slip that causes
oxide film rapture hence initiating and propagating IG cracks. The radiation-induced

8. segregation alters SFE leading to the enhancement or diminishing of the planarity slip, thus
causing changes at the grain boundary. The RIS is found to be dependent on factors like GB
structure, the composition of the alloy, and the dose of irradiation. The structure of GB has
the highest capability for suppression of segregation to GBs under irradiation. The
dependency of loop size and number density on conditions of irradiations and alloying
elements was contradicted by a study conducted on chromium and nickel. The results
obtained indicated no observed correlation between alloy chemistry and loop size or
density. The primary cause for ISACC has not been identified as it is depended on both the
way deformation of a material occurs and the state of the material that is irradiated. The
rapture of oxide films near the GB is a potential ISACC mechanism.
Bibliography
(1) Hojná, A. Irradiation-Assisted Stress Corrosion Cracking and Impact on Life Extension.
CORROSION 2013, 69 (10), 964–974. https://doi.org/10.5006/0803.
(2) Was, G. S.; Wharry, J. P.; Frisbie, B.; Wirth, B. D.; Morgan, D.; Tucker, J. D.; Allen, T. R.
Assessment of Radiation-Induced Segregation Mechanisms in Austenitic and Ferritic–
Martensitic Alloys. Journal of Nuclear Materials 2011, 411 (1-3), 41–50.
https://doi.org/10.1016/j.jnucmat.2011.01.031.
(3) Was, G. S.; Busby, J. T. Role of Irradiated Microstructure and Microchemistry in
Irradiation-Assisted Stress Corrosion Cracking. Philosophical Magazine 2005, 85 (4-7), 443–
465. https://doi.org/10.1080/02678370412331320224.
(4) Jiao, Z.; Was, G. S. The Role of Irradiated Microstructure in the Localized Deformation of
Austenitic Stainless Steels. Journal of Nuclear Materials 2010, 407 (1), 34–43.
https://doi.org/10.1016/j.jnucmat.2010.07.006.
(5) Tan, L.; Stoller, R. E.; Field, K. G.; Yang, Y.; Nam, H.; Morgan, D.; Wirth, B. D.; Gussev, M. N.;
Busby, J. T. Microstructural Evolution of Type 304 and 316 Stainless Steels under Neutron
Irradiation at LWR Relevant Conditions. JOM 2015, 68 (2), 517–529.
https://doi.org/10.1007/s11837-015-1753-5.
(6) Xu, C.; Zhang, L.; Qian, W.; Mei, J.; Liu, X. The Studies of Irradiation Hardening of Stainless
Steel Reactor Internals under Proton and Xenon Irradiation. Nuclear Engineering and
Technology 2016, 48 (3), 758–764. https://doi.org/10.1016/j.net.2016.01.007.
(7) Jiao, Z.; Was, G. S. Impact of Localized Deformation on IASCC in Austenitic Stainless
Steels. Journal of Nuclear Materials 2011, 408 (3), 246–256.
https://doi.org/10.1016/j.jnucmat.2010.10.087.
(8) Raquet, O.; Herms, E.; Vaillant, F.; Couvant, T.; Boursier, J.-M. Scc of cold-worked

9. austenitic stainless steels in pwr conditions
https://www.researchgate.net/publication/266507825_SCC_of_cold-
worked_austenitic_stainless_steels_in_PWR_conditions.
(9) Onchi, T.; Dohi, K.; Soneda, N.; Navas, M.; Castaño, M. L. Mechanism of Irradiation
Assisted Stress Corrosion Crack Initiation in Thermally Sensitized 304 Stainless Steel.
Journal of Nuclear Materials 2005, 340 (2-3), 219–236.
https://doi.org/10.1016/j.jnucmat.2004.11.012.
(10) Villacampa, I.; Chen, J. C.; Spätig, P.; Seifert, H. P.; Duval, F. Helium Effects on IASCC
Susceptibility in As-Implanted Solution Annealed, Cold-Worked and Post-Implantation
Annealed 316L Steel. Corrosion Engineering, Science and Technology 2017, 52 (8), 567–
577. https://doi.org/10.1080/1478422x.2017.1323709.
(11) Chung, H. M.; Ruther, W. E.; Sanecki, J. E.; Hins, A.; Zaluzec, N. J.; Kassner, T. F.
Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels: Recent
Progress and New Approaches. Journal of Nuclear Materials 1996, 239, 61–79.
https://doi.org/10.1016/s0022-3115(96)00677-0.
(12) Du, D.; Sun, K.; Was, G. S. IASCC of Neutron Irradiated 316 Stainless Steel to 125 Dpa.
Materials Characterization 2021, 173, 110897.
https://doi.org/10.1016/j.matchar.2021.110897.