E-Journal of Advanced Maintenance Vol.4 No.2 (2012) 36-56
Japan Society of Maintenology

Proactive Material Degradation Research Subjects
for Light Water Reactors
Tetsuo SHOJI1,*, Yoichi TAKEDA1, , Jiro KUNIYA1, Peter FORD2 and Peter SCOTT3
1

Tohoku University, 6-6-01 Aoba, Aramaki, Aoba-ku, Sendai 980-8579, Japan
Consultant, 19 Nott Road, Rexford, NY 12148, USA
3
Consultant, 4 Résidence Les Vergers, 78590 Noisy Le Roi, France
2

ABSTRACT
Predictive and preventive maintenance technologies are increasingly of importance for the long term operation(LTO) of Light Water Reactor (LWR) plants. In order for the LTO to be successful, it is essential that aging
degradation phenomena should be properly managed by using adequate maintenance programs based onforeseeing the aging phenomena and evaluating their rates of development. In this paper, topics that are judged
necessary for managing aging degradation phenomenon in LWR structural materials and which were discussed
in the Proactive Aging Management Experts' Panel Meeting of Tohoku-Hokkaido Research Cluster of NISA
Project are introduced
KEYWORDS

ARTICLE INFORMATION

Proactive materials degradation assessment, Proactive materials degradation
management, LWR, BWR, PWR, Plant Aging, Long Term Operation, Stress
Corrosion Cracking, Flow Assisted Corrosion

Article history:
Received 25, Nov 2011
Accepted 4, June 2012

1. Introduction
A project on Proactive Materials Degradation Management (PMDM) has been carried out at
Fracture and Reliability Research Institute, Tohoku University for four years, as a part of the Aging
Management Program of the Nuclear Industries Safety Agency (NISA) that was started in 2007. The
objectives of this project are to evaluate potential and complex degradation phenomena and theirmechanisms in order to identify future risks of component aging in nuclear power plants. The
following items are of particular concern in this project. (a) Investigation of potential materials ageing
phenomena and corresponding plant issues, and (b) Investigation of the effectiveness of evaluation
techniques concerning potential aging phenomena.
Similar to the PMDM approach of the US Nuclear Regulatory Commission (USNRC) [1], a
Phenomena Identification and Ranking Table (PIRT) approach was adopted. For this purpose, a panelof international experts on materials degradation in light water reactor (LWR) were nominated andassigned to this task. One major difference between the present PMDM project and the USNRC
PMDM is that the preferred approach in the current PMDM project is based upon a more fundamental
understanding of the scientific bases of materials ageing. Two approaches are considered to be an
essential for proactive aging management. One is a deductive scientifically based method and anotheris an intuitive one based on a careful analysis of operating experiences. In particularly, the deductive
method based on a fundamental scientific understanding of material degradation is paramount for the
management of possible latent materials degradation processes which have not yet become obvious inoperating plants.
*

Corresponding author, E-mail: tshoji@rift.mech.tohoku.ac.jp

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T. Shoji, et al./ Proactive Material Degradation Research Subjects for LWR

Many of these fundamental and managerial issues impact on current concerns associated with, forexample, flow accelerated corrosion, stress corrosion cracking (SCC) of nickel-base alloys in primarysystems of pressurizes water reactors (PWRs), and irradiation assisted SCC (IASCCC) of stainlesssteels in core structures of boiling water reactors (BWRs) that were initially prioritized in 2007. Later,in 2010, proactive management issues associated with materials aging in LWRs were discussed interms of suggested research topics that should be undertaken in either the short or long term. Based onthese discussions prioritized lists of medium and long term research projects were established for both
PWRs and BWRs.
In this paper, the research subjects to be considered for the aging degradation phenomena in LWR
structural materials that were discussed at the Proactive Aging Management Experts' Panel Meeting of
the NISA Project are introduced.

2. Research Subjects for Basic Research on Environmentally Assisted
Degradation of BWR Materials.
2.1 Research subjects for shorter term issues to be completed in ~5 years in BWRs
2.1.1 Definition of the Metal/Environment Interface and Precursor Events[Background]
It has long been recognized that the environmentally-assisted cracking susceptibility of structuralsteels and nickel base alloys in BWRs is governed by a significant number of material, stress andenvironmental parameters and that there are secondary interactions between these parameters.

For

instance, both the heat input during welding and the fast neutron flux/fluence will affect not only thegrain boundary composition in stainless steels, but also the residual stress profile adjacent to welds
and the corrosion potential.
It is obvious that, if the kinetics of EAC are to be predicted, then the above materials, environmentand stress inputs must be adequately defined, as well as their interactions.

Although such systemdefinitions have been developed for many of the critical parameters (for example, the extent ofchromium depletion adjacent to grain boundary carbides as a function of heat treatment), this is not
the case for all parameters and their interactions.
[Scope]The scope of this particular RFP focuses on the quantification of critical items (e.g. corrosion
potential, anionic activity, stress distribution, material composition) that are central to the predictionof environmentally-assisted degradation in stainless steel, nickel-base alloy and carbon or low alloysteel BWR components.

Such quantification involves a basic understanding of, for example, thedislocation morphology and plasticity at crack tips, and the thermodynamics and kinetics ofelectrochemical reactions and metallurgical transformations. However for practical purposes thesesystem conditions must be redefined in terms of measureable or calculable parameters such as stress,
strain, heat treatment, water purity, component geometry, etc.. This is often difficult. For instance, theeffect of cold work on the cracking susceptibility of stainless steels and nickel-base alloys in BWRs
may be governed by a wide range of interacting physical metallurgical phenomena, and it would bevery difficult to formulate these interactions in terms of a fundamentally important parameter such as
the strain rate at a crack tip. At the very least, however, it should be possible to “bin” combinations ofthese engineering conditions that have a similar effect on the fundamentals of crack initiation and37

T. Shoji, et al./ Proactive Material Degradation Research Subjects for LWR

propagation.
2.1.2 Quantification of Crack Initiation for Unirradiated Structural Alloys in BWRs.
[Background]
There are various definitions of "crack initiation" ranging from the initiation of isolated
micro-cracks at pits or sites of inter-granular corrosion, to the detection of deeper cracks by a variety
of non-destructive testing procedures conducted on the reactor component surface. Theinterrelationship between these definitions is illustrated in Figure 1. The definition of crack initiation
adopted in this RFP is the time, tO , for a dominant crack to be formed by the coalescence of separated
micro-cracks. A fairly extensive number of investigations indicate that, for stainless steels, nickel-basealloys and carbon/low alloy steels in BWR environments, such coalescence is completed when thedominant crack depth, ao , is in the range 50-300µm . The subsequent crack propagation rate isapproximately the same (at the same nominal stress intensity factor) as that associated withpropagation rate studies

on very much deeper cracks, well in excess of the approx.1000 µm

associated with crack detection.

Thus, there is an analytical pathway to quantifying lifetime

Crack Depthpredictions for BWR components.

ad “1000” µm

Design limitation

‐1σ

mean

+1σ

td

tcritical

Time

Detected crack

Coalesced crack
a o “50‐300”µm

t0

Figure 1. Schematic crack depth vs. time relationships associated with the definitions of crack"initiation" and propagation
It is widely accepted that many of the sub-phenomena that lead to a dominant coalesced crack (suchas, inter-granular attack and pitting) are stochastic in nature, and hence, as illustrated in Figure 1, there
will be an expected distribution in to values associated with the specific ao value appropriate for agiven alloy environment combination. Moreover this dispersion is carried through to evaluations ofthe "engineering initiation" time, td, and to the time when the first crack reaches some design criterion.
There have been numerous hypotheses advanced to explain the variations in "initiation time", andthese have ranged from the relatively simple explanation of crack initiation due to inter-granular
attack of a chromium-depleted grain boundary in sensitized stainless steel, to more involved
hypotheses to explain crack initiation on the cold worked surface of non-sensitized stainless steels.
These latter hypotheses have concentrated primarily on wrought stainless steels, rather thannickel-base alloys, and have centered on the role of the oxide (compositionand structure) andcomplex changes in metal microstructure immediately below the metal oxide interface (e.g.

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T. Shoji, et al./ Proactive Material Degradation Research Subjects for LWR

recrystallized nano-dimension grain size, slip bands/twins), and strain localization at grain boundaries.
Hypotheses to explain the early micro-crack initiation and coalescence for carbon and low alloysteels have followed the much earlier work in lower temperature/concentrated environments, which
focused on micro-crack initiation in pits and the spatial aspects of the pits that enabled crack
coalescence. Preliminary observations on such ferritic alloys under fatigue loading in the higher
temperature BWR environments indicate that a similar phenomenon is occurring, provided the pits
have been initiated at lower temperatures.
has yet to be fully quantified.

However this hypothesis for carbon and low alloy steels

A more recent hypothesis has been advanced for cold worked low

alloy steels based on the effect of absorbed hydrogen (produced by flow accelerated corrosion at
adjacent geometrical discontinuities) leading to creep crack initiation and growth. This is a distinct
possibility for PWR secondary side components operating at higher temperatures, and has yet to beevaluated for BWR components.
[Scope]
The aim of this research subject is to (a) directly measure the values (and dispersion) of theparameters a0 and to shown schematically in Figure 1, and (b) to correlate these observations withmicro-structural changes in the surface region.
2.1.3 Development of Qualified Mechanisms-Based Algorithms for EAC Propagation in
Unirradiated Structural Alloys.
[Background]
There have been extensive reported investigations over the last 30 years to develop prediction
models for environmentally-assisted crack propagation (i.e. stress corrosion cracking and corrosion
fatigue) in structural alloys in the BWR RCS. These developments were motivated by a need tounderstand (and potentially simplify) the interpretation of the extremely complex material,
environment and stress interactions. Most attention in this regard has been given to unirradiatedstainless and carbon/low alloy steels with further preliminary developments for nickel-base alloys,
and irradiated stainless steels.
The mechanism of crack propagation that has been studied the most (primarily at General Electric)is the slip oxidation model. The central hypothesis is that crack advance may be directly related to theoxidation kinetics (i.e. dissolution and oxide formation) at the straining crack tip where thethermodynamically stable oxide is being periodically ruptured. Under such conditions the crack
propagation rate (da/dt ) may be related to a formulation of the general form;
da/dt = A (dε/dt) n
where

A = a constant that includes the value n, the fracture strain of the oxide at the crack tip
and Faradays constantdε/dt = the crack tip strain rate, which controls the oxide rupture frequency and which
may be empirically formulated in terms of calculable engineering parameters,such as stress intensity factor, etc.
andn = parameter that quantifies the oxidation rate on the bared crack tip created by therupture of the crack tip oxide. This value is a function of the degree ofchromium depletion at the grain boundary the corrosion potential at the crack
mouth and the water conductivity.

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Developments at the Fracture Research Institute (FRI), Tohoku University have expanded on theseearlier developments at General Electric in terms of the details of the oxidation process and, inparticular, the formulation of the crack tip strain rate at an advancing crack tip.

The developmentson the former topic are still in the formative state, but the developments in the latter topic have given
insight into the reasons for the enhanced cracking susceptibility due to cold work observed in
non-sensitized alloys and under more reducing environments.
Very similar predictions have been made using the slip oxidation hypothesis for stress corrosion
crack propagation in carbon and low alloy steels and these have been accepted by the USNRC forsupporting disposition relationships for pressure vessel steel[Scope]In spite of these promising agreements between observation and theory, there remain issues that put
to question some of the basic science assumptions behind the General Electric and FRI prediction
models. These need to be addressed, and in particular;(1)The nano-meter scale of the crack tip opening and the effect that this has on the mass transport
controlled oxidation reactions at the crack tip. The current crack propagation rate prediction
algorithms are based on reaction rates measured in simulated crack tips with unrestricted fluid
access.
(2)The (re)formulation of the crack tip strain rate to take into account the yield stress in the bulk
metal. Such formulations have been developed using strictly empirical (General Electric) andanalytical (FRI) approaches. The assumption in this development is that it does not matter how theyield stress is changed (e.g. cold work, irradiation damage, heat treatment). This seems rather
over-simplistic and needs more justification.
(3)The formulation of the crack tip strain rate that takes into account localized microstructural andplasticity phenomena . Such phenomena include;(a) the role of stress intensity factor gradients (i.e. dK/da). Positive values of dK/da are observed togive significant increases in the crack propagation rate, which may be relevant during the growthof short cracks.
(b) the role of stress/strain transients on the creep rate(c) the role of absorbed hydrogen on creep rates.
(d) the role of dislocation morphology and (microscopic) strain localization. The dislocation
morphology and how it is changed by such factors as the stacking fault energy, coherent
precipitates, grain boundary sliding, etc. has long been recognized as being relevant to stresscorrosion cracking.

Its effect is understood qualitatively in terms of changes in the crack tip

strain rate, but there is, as yet, no quantitative correlation.
It is possible that the resolution of these basic science issues will uncover situations where there arecompensating errors in the current working hypothesis.

However, it is likely that the basic

hypothesis is correct in principle, since it gives a reasonable prediction of observed SCC propagationsusceptibilities for stainless steels (and carbon/low alloy steels) over a wide range of systemparameters. If this presumption is correct then this hypothesis should be expanded to cover theprediction of crack propagation in;Stainless steels in irradiation environments (e.g. core internals).

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T. Shoji, et al./ Proactive Material Degradation Research Subjects for LWRThe prediction of crack propagation under changing load and environmental conditions.



Nickel base alloys (e.g. 600,182,82) that are used in safe ends, attachment welds, access holecovers.

Stress corrosion crack propagation of irradiated low alloy steels.

2.2 Research subjects for Long Term Issues to be completed in 5-10 years in BWRs.
2.2.1 Prediction of Stress Corrosion Cracking of Stainless Steels in BWR Core Internals[Background]
The SCC susceptibility of the irradiated components in BWRs responds to the various system (i.e.
material, stress and environmental) parameters in much the same manner as has been observed for
unirradiated stainless steels.

For instance, the cracking susceptibility in both cases is dependent on

the tensile stress, corrosion potential, anionic impurity concentration and the grain boundary
chemistry. Thus, it is reasonable to propose that IASCC is not a new mechanism of cracking, but ismerely mirroring irradiation-induced changes to the rate-controlling parameters in the various phases
in crack development described earlier in sections 3.1.2. and 3.1.3. Such irradiation-induced changes
would be to the corrosion potential, tensile residual stress, grain boundary composition and yieldstress. Those changes that are driven by neutron flux (i.e. corrosion potential) will have an effectimmediately on the cracking susceptibility, whereas other fluence-driven changes (such as to material
composition, stress relaxation and yield strength) will have an effect over an extended time, albeit
over periods that are relevant to current BWR operating times.
It has been hypothesized, therefore, that the slip-oxidation mechanism for crack propagation is therelevant cracking model for irradiated stainless steels in BWRs. The task, therefore, is to identify theroles of neutron flux, fluence and gamma irradiation on the corrosion potential, applied stress, theyield stress and the grain boundary chemistry.
Programmatically, the specific effect of irradiation on the fundamental parameters that control
crack propagation in stainless steels may be quantified via consideration of the following;

Corrosion potential and how this changes with radiation flux.



Irradiation-induced changes in grain boundary composition.



Irradiation-induced hardening and, for displacement loaded structures, stress relaxation.

The results of preliminary investigations into such irradiation-induced alterations to theslip-oxidation prediction model for crack propagation stainless steels have been reported. For instance,
it is apparent that the crack depth is predicted to change with time (fluence), in a manner that reflects
the time-dependent reduction in weld residual stress (which decreases the cracking susceptibility) andthe increase in irradiation-induced grain boundary chromium depletion (which increases the cracking
susceptibility). In this case the total grain boundary chromium depletion is attributed to any initial
thermally-induced chromium depletion at the grain boundary due to the welding process plus the
additional depletion that occurs with fluence at that particular point in the shroud.

In addition, theyield stress will increase with fluence and will increase the cracking susceptibility due to its effect on

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T. Shoji, et al./ Proactive Material Degradation Research Subjects for LWR

the crack tip strain rate. Thus, it is apparent that the crack growth rate is not constant but will vary,mirroring these competing effects of stress relaxation, increasing yield stress and grain boundary
sensitization, etc.. (Note that this observation immediately imposes a fluence/time limitation on thequantification of crack propagation rate vs. stress intensity factor disposition relationships)
It is apparent from the discussion above that there is a basis for predicting crack propagation ofirradiated stainless steels. However, exactly as with concerns about IASCC in PWR corecomponents, there are potentially extra issues at irradiation fluences circa5 to 10dpa whichmight be encountered under BWR lives of 60-80 years and which are not experienced atlower fluences. As an example, there is increasing evidence that silicon segregation may be ofimportance. Silicon enrichment at the grain boundaries increases with fluence and ispotentially of concern because many stainless steels containing 0.5 – 1% Si can enrich to>5% at the grain boundary. Indeed, since the compositional measurements are generally
made by analytical electron microscopy with a 1 – 2 nm beam size, the actual Si
concentration at the grain boundary might approach 50 atomic percent. Crack growth ratemeasurements on stainless steels with elevated Si levels (e.g., 1.5 – 5% Si) show high growth
rates and limited or no effect of stress intensity factor and corrosion potential. Thisobservation may help to explain the loss of the benefit of lowering the corrosion potential athigh fluence in some stainless steels, especially since Si enrichment appears to continue after
Cr depletion saturates. Unfortunately, there is no quantitative life prediction capability forthis effect at this time.
[Scope]The basic science issues, (apart from the more "applied science" concerns associated withthe validation of radiolysis/ corrosion potential models), relate to the definition of the grainboundary properties and how these alter with fast neutron fluence. Concerns centre aroundthe fundamental understanding of the role of silicon segregation, hardening, helium bubble
and late blooming phase generation on IASCC. This understanding should be linked to thedevelopment of the IASCC propagation model described above coupled with developing
better models of microstructural development under irradiation. Such microstructuralmodels should take into account interstitial diffusion phenomena and the effects of neutron
spectrum on nuclear transformation reactions leading to helium and hydrogen generation.
Emphasis in EAC testing should be on both initiation and crack growth testing with attention
in the former case as an extension of Section 3.1.2 on the role of irradiation on surface
hardening, oxide formation and degradation prior to crack formation. The role of channel
formation during deformation and interaction with grain boundaries should also be
investigated, with reference to the definition of the crack tip strain rate algorithm.
2.2.2 Life-Prediction Capability for EAC of Structural Alloys in BWR Components[Background]
Various approaches have been taken in the past to predict the life of a component subject toenvironmentally-assisted cracking. These have included; (a) analyses of failure times for specificcomponents, (b) statistical approaches for evaluating the role of the different stress, material andenvironmental "drivers", (c) the evolution of empirically-based prediction algorithms and, (d) in a few

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cases, the development of life-prediction algorithms that are based on an understanding of the more
fundamental phenomena occurring at the metal/environment interface. All of these approaches have
limitations either because they do not recognize the chronology of physical events, such as a precursor
time, microcrack initiation at grain boundaries or pits, microcrack propagation and coalescence and,
finally, macrocrack propagation or, if they do, the resulting prediction algorithms do not adequately
quantify in local and global terms the specifics of these phenomena.

A widely reported approach, which might act as a basis for future lifetime predictions, isthat taken for predicting the life of stainless steel piping. In this case it was assumed that an
intrinsic defect of depth 50µm was present at the time of reactor commissioning; thisassumption is in line with the conservative observation. It was further assumed that theresultant growth of that crack was predicted via the slip oxidation model (section 3.1.3.)
using the relevant stress intensity/crack depth relationship for the classification of the piping
system and the relevant water conductivity, corrosion potential and sensitization. The range incrack depths noted at one reactor (operating under very (by modern standards) impure waterconditions are predicted, with the maximum crack depths corresponding to the predictions forthe maximum residual stress for that 28" recirculation piping system. By comparison, as
predicted, no cracks were observed (i.e the crack depths did not exceed the NDT detection
limit) during the examination period. As expected, the time for the cracks to reach adetectable depth corresponding to a quarter wall penetration was a function of the water
purity, with the precise dependency being a function of the residual stress profile and thedegree of sensitization. Such a prediction capability leads to the ability to predict futuredegradation kinetics as a function of, for instance, the planned operating conditions.
[Scope]The objective is to build on the conclusions arising from Sections 3.1.1, 3.1.2, and 3.1.3, and todevelop life prediction algorithms for stainless steels, nickel-base alloys and low alloy steels that are
subject to environmentally-assisted cracking in BWRs under a variety of operating modes. The
achievement of this objective requires;The experimental determination of the mean values of a0 and t0 in Figure1 that relate to theachievement of a dominant coalesced crack. There will be definable aleatory uncertainties
associated with these parameters, which are inputs to;The propagation rate algorithm for the specific alloy/environment that is relevant to a given
BWR component. The epistemic uncertainties associated with this sub-objective relate toquestions concerning the completeness of the prediction algorithm and the uncertainty in the
inputs to that algorithm.

It is emphasized that this overall task can start immediately, using currently available models andmodel inputs (or assumed ranges of these inputs).

Success is therefore quantifiable in terms ofpredicting the mean and dispersion of laboratory and/or plant data.
2.2.3 Development of an in-situ monitoring /diagnostics and a risk-informed management
capability for BWR components subject to EAC.
[Background]

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Two developments are technically achievable based on the proposed work described in the previous
sections, both of which are compatible with proactive management of materials degradation of
structural materials in BWRs. These interelated developments are; (a) a real-time, in-situ monitoring
capability and (b) a risk-informed management capability.
[Real-time , in-situ monitoring capabilities]
Real-time, in-situ monitoring capabilities have been developed for other industries including thefossil power generation, aircraft and transport industries. There is no reason in principle that theycannot be developed and used to manage materials degradation issues in BWRs.
This capability requires;


System definition “tools", associated with; radiolysis and corrosion potential models; models
associated with the definition of the material condition that are dependent on the fabrication
process; etc.



System definition monitors placed in accessible positions that act as calibration points to themodel. Such monitors include corrosion potential and crack growth rate specimens.

 Life prediction models for specific material / environment / degradation mode combinations.
Elements of all these requirements are already available, or are in development as described in theprevious sections.
The use of such a monitoring and diagnostics capability leads to;

The instantaneous definition of changes in the "frequency" input to a risk analysis that might
be associated with , for instance, changes in material degradation due to unplanned changes in
the reactor operating conditions



Assessments of "factors of improvement" (FOI) relating to potential changes in theenvironment.

 Prioritization of inspection actions based on current degradation predictions
Although preliminary versions of such an "M&D" integrated system may be tested now, there are
significant obstacles before "commercial" application would be possible.

These include; The qualification of radiolysis and corrosion potential models, with appropriate global to local
transition capabilities.
 The development of long life, reliable corrosion potential monitors.
 The development of non-destructive, in-situ, residual stress monitors and methods to quantify
the local strain condition.
[Risk-informed management capability that address time-dependent materials degradation]
Risk-informed safety analyses have been in place primarily in the US since the 1980s for regulationand operational decisions (e.g. inspection, maintenance, pressurized thermal shock). Thesemethodologies, which combine traditional deterministic safety analyses with probabilistic risk
assessment, need to be extended to provide a risk-informed safety margin characterization that takes
into account materials degradation that occurs over time.

3. Research Subjects for Basic Research on Environmentally Assisted
Degradation of PWR Materials
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T. Shoji, et al./ Proactive Material Degradation Research Subjects for LWR

3.1 Research subjects for shorter term issues to be completed in ~5 years in PWRs
3.1.1 Stress Corrosion Crack Initiation Phenomena, including the Effect of Surface
Stress/Strain, Residual Stress, Microstructure and Strain Localization[Background]
Crack initiation is inherently difficult to study because of the stochastic nature of the processes thatusually requires many specimens to be tested under identical conditions in order to determine themean (and standard deviation) of the response. Nevertheless, with proper attention to specimen
preparation with the objective of ensuring that all known observable surface characterization
parameters are held constant, significant progress is considered attainable even when testing only
modest numbers of replicated specimens, say ~6 to determine mean behavior with a reasonable levelof statistical confidence.
The two SCC phenomena where most attention is needed in relation to characterizing initiation
response are Primary Water Stress Corrosion Cracking (PWSCC) of nickel base alloys, particularly
Alloy 600 and the weld metal Alloys 132/182, and Inter-granular Stress Corrosion Cracking (IGSCC)
of cold worked stainless steels, also in nominal PWR primary Water conditions. However, in thesecond case, the role of water chemistry transients on the process is an additional variable requiring
clarification. In addition to these two principal topics, SCC of carbon and low alloy steels (C&LAS)
and external surface SCC (mainly of austenitic stainless steels) are additional topics meriting attention
in the context of long term plant aging.
[Scope]Existing research results have clearly revealed the importance of surface condition in relation to theprobability of PWSCC initiation in nickel base alloys in addition to the usual parameters of stress,
corrosion potential, temperature and the inherent variability in susceptibility of individual heats ofmaterial. However, there is dearth of quantitative information relating to the characterization of
surface condition and effects on crack initiation probability. Among the parameters that could beassessed are surface residual stress, near-surface microstructure including dislocation density, grainboundary carbide coverage, grain boundary coherency, and strain localization.
The position with SCC initiation in cold worked stainless steels is less clear in that almost no work
has been published on IGSCC initiation in PWR primary water other than via the dubious means ofslow strain rate tests. Essentially the same parameters as those suggested above for nickel base alloys
could be studied with the addition of the effect of transient oxidizing conditions of varying durationthat typically occur at lower temperatures than normal operating temperature and are strongly
suspected as being implicated in IGSCC initiation. Periodic slow dynamic straining is anotherparameter requiring fundamental assessment.
In the case of C&LAS, considerable attention has been focused in the past on the influence oftransient oxidizing conditions and of transient dynamic straining on SCC initiation (and propagation).
However, some recent plant events have drawn attention to the possibility of crack initiation in cold
worked materials either at low corrosion potentials or even on external pipe surfaces, possibly due tothe flux of hydrogen from internal surface corrosion and flow assisted corrosion. On the other hand,external surface SCC of austenitic stainless steels caused by chloride contamination is a very wellcharacterized phenomenon and it is difficult to see what further fundamental work is required.

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3.1.2 Development of Qualified Mechanisms-Based Life Time Models for PWSCC
Propagation in Nickel Base Alloys[Background]
The challenge here is the absence of a consensus on the mechanism of PWSCC propagation as wellas an absence of mechanistic explanations for the considerable variability that is observed between
different heats of Alloy 600 or different Alloy 132/182 weld metal deposits. Some considerable
advances have been made by the Tohoku Fracture group in the application of modern solutions for the
stress/strain fields in front of crack tips combined with various hypotheses for the mechanism bywhich cracks may advance in PWR primary water environments. The latter include the classicalapplication of Faraday’s law to metal removal by dissolution/oxidation at a crack tip or embrittlement
mechanisms involving oxygen. Similar approaches to hydrogen embrittlement effects have been
explored by M. Hall and colleagues.
[Scope]In the shorter term, it would appear preferable to examine modeling of a single system such as
PWSCC growth in Alloy 600 and endeavor to distinguish between the dominance of different
potential mechanisms both theoretically and experimentally. The objective, therefore, is to predict
quantitatively the effects of all the major known variables of corrosion potential (i.e. hydrogenconcentration in PWR primary water), stress, temperature, cold work (including strain path), andcarbide morphology. From the existing models incorporating the three proposed rate determining
processes at crack tips involving oxidation rates as given by Faraday’s law, oxygen diffusion in thecrack tip oxide and/or metal, and hydrogen diffusion/accumulation in the crack tip plastic zone, thecrack growth rate as a function of KI should be forecast and then tested methodically in the laboratory.
The influence of corrosion potential and microstructure are perhaps the most revealing in terms oflikely observable differences between the various mechanistic hypotheses for PWSCC growth.
Another important issue is the behavior of long term aged materials, which could include materials
removed from operating plants during the course of component replacements.
3.1.3 Characterization of Weld Metals, Dilution at interfaces, and Heat Affected Zones[Background]This topic concerns primarily dissimilar metal welds between RPV low alloy steels and stainlesssteel safe ends (both wrought and CASS) where the weld metals may be Alloys 132/182, 82, 152 or
52 or Type 309/308 stainless steel. The concern is for the toughness and stress corrosion resistance ofthe various metallurgical discontinuities within dissimilar metal welds and how these vary withwelding practices. In particular, the residual stresses and microstructures of interfaces between weld
metals and base metals require characterization and ultimately correlation with toughness andresistance to SCC, particularly in PWR primary water in the current context.
[Scope]For this project, an alliance with a qualified fabricator of nuclear components is essential in order tocover the correct ranges of welding methods, welding parameters, weld profiles and weld constraint.
Subsequent metallurgical characterization is by classical optical and electron microscopy coupledwith determination of residual stress profiles by X-ray diffraction and hole drilling techniques. An
examination of the risks of hot cracking of the weld metal and its dependence on weld metal

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composition and other welding parameters could be studied concurrently.
It would also be desirable to correlate this metallurgical information with measurements of fracture
toughness of the various interfaces and their resistance to PWSCC. The latter could be achieved byfour point bend specimen exposures and crack propagation rate measurements in simulated PWR
primary water. The optimum temperature for the maximum acceleration of testing would have to be
determined as this may not necessarily be the same for all the materials used for dissimilar metal
welds.
3.1.4 Strain Localization, Strain History and Relationship to Cold Work, Microstructure andCompositional Banding[Background]
Cold work, particularly in one plane due to rolling, is known to provoke PWSCC susceptibility inwrought thick section Alloy 690 and austenitic stainless steels that are otherwise extremely resistant tocracking. The issue is confused by discrepancies between different reports of the importance or not of
carbide banding and its interaction with cold work, especially in wrought Alloy 690 plates. There is,therefore, a need to systematically examine the relationship between banding, cold work in the sameplane as the banding, and PWSCC propagation rates in that plane. In the case of Alloy 690 there isalso a question concerning the differences to be expected for relatively fine grained, steam generator
tubing in the mill annealed and thermally treated condition compared to wrought products.
[Scope]At first sight it would not seem feasible to procure one plate of Alloy 690 with pronounced carbide
banding and process it to remove the banding but feasibility could be examined in more detail. Failingthat, two separate plates should be procured, with and without pronounced carbide banding, andcharacterized by classical optical and electron microscopy. In addition, strain localization at various
microstructural features should be quantified using EBSD techniques.
Further cold work up to typically 40% cold reduction should be performed on the two types of
microstructure and from which CT specimens are subsequently manufactured with the crack plane atdifferent depths through the half-thickness. The aim will then be to carry out PWSCC growth ratemeasurements and compare the effect of different microstructures.
Material from Alloy 690 steam generator tubing in the mill annealed and thermally treated condition
should also be compared in crack growth tests and combined with microstructural characterizationand comparisons.
3.2 Research subjects for Long Term Issues to be completed in 5-10 years in PWRs.
3.2.1 Effects of Irradiation Flux and Fluence on Stainless Steels and Nickel Base Alloys andthe Effects on EAC[Background]
Changes in elemental composition adjacent to grain boundaries of austenitic alloys, particularly
stainless steels, under neutron irradiation have been well characterized and documented. They
typically include depletion of chromium, iron and molybdenum in a band about 5 to 10 nm wide
either side of the grain boundary and corresponding enrichment in nickel and silicon. The evolution in
grain boundary composition is observed between neutron doses of 1 to 10 dpa and tends to saturate at47

T. Shoji, et al./ Proactive Material Degradation Research Subjects for LWR

neutron doses of ~10 dpa.
Modeling of the vacancy driven inverse Kirkendall mechanism of atom migration under acontinuous neutron flux is also well advanced for the so-called over-size atoms like chromium, ironand molybdenum. However, the corresponding grain boundary enrichment mechanism of theso-called under-sized atoms such as nickel and silicon that probably proceeds by an interstitial jumpmechanism is much less well developed and there is no known theoretical model in the literature thathas been calibrated with the experimental observations.
There are also concerns over the possible differences between fast reactor and thermal reactor
irradiated microstructures due to neutron spectrum and flux effects, in particular the generation oflarge quantities of helium and molecular hydrogen trapped in nanometer scale bubbles observed inthermal reactor irradiated materials and the extent to which grain boundaries may be preferential sites
for bubble formation. There is no clear picture of how these processes affect EAC (and mechanicalproperties) of, in particular, PWR core structural materials. The effect of late blooming phases athigher irradiation doses is another poorly characterized phenomenon in this respect.
Depletion of chromium at grain boundaries is particularly detrimental for IASCC in the moreoxidizing environment typical of BWRs on Normal Water Chemistry without any deliberately addedhydrogen. However, PWR operating experience also shows IASCC susceptibility of irradiatedaustenitic stainless steels in the same range of neutron doses even though chromium depletion is of no
particular consequence at the low corrosion potentials typical of PWR primary circuit materials and in
the aqueous phase of BWRs on hydrogen water chemistry. In PWRs, IASCC susceptibility may
continue to increase with even higher neutron doses above 10 dpa characteristic of some parts of
PWR internal core support structures and suggestions have been made that silicon segregation inparticular may be detrimental. Thus, the mechanism of IASCC in hydrogenated reactor coolants like
PWR primary water and BWR hydrogen water chemistry is not properly elucidated. More generally,
the interaction between grain boundary elemental segregation and matrix hardening induced byneutron irradiation plus the formation of nanometer scale helium gas bubbles and precipitation of ‘late
blooming’ phases in highly irradiated materials typical of PWR core supports and their effect on
IASCC susceptibility remain to be elucidated.
Research work on highly irradiated and activated austenitic alloys like stainless steels and nickel
base alloys requires a particularly long term commitment of resources both material and financial. It isalso essential for fundamental studies that a university establishes a working alliance with a properly
equipped hot laboratory that can devote the resources and manpower required to develop andcommission the in-cell equipment required. Links to the fusion reactor community may also be
usefully developed, in particular to take advantage of developments in techniques such as co-injectionof energetic ionized hydrogen and helium.
[Scope]The lines of approach suggested at this stage of the development of the knowledge of IASCC in
order to improve fundamental understanding are to examine the role of silicon segregation, hardening,
helium bubble and late blooming phase generation on IASCC. This should be linked to thedevelopment of an IASCC model based, if possible, on a mechanistic foundation and coupled todeveloping better models of micro-structural development under irradiation taking into accountinterstitial diffusion phenomena and the effects of neutron spectrum on nuclear transformation48

T. Shoji, et al./ Proactive Material Degradation Research Subjects for LWR

reactions leading to helium and hydrogen generation. Emphasis in EAC testing should be on both
initiation and crack growth testing with attention in the former case on the role of irradiation on
surface hardening, oxide formation and degradation prior to crack formation. The role of channel
formation during deformation and interaction with grain boundaries should also be investigated.
3.2.2 Modeling and Validation of Residual Stress/Strain Profiles in Complex Welded
Geometries and how these may Change with Neutron Fluence[Background]In many respects, this topic is more an engineering problem of determining residual stresses incomplex three dimensional engineering components and how these re-arrange due to the action ofirradiation creep and other physical changes induced by neutron irradiation. However, an importantelement of such assessments concerns the details of weld residual stress/strain profiles on a more local
scale in welded joints and relaxation by irradiation creep. It is the irradiation creep aspect of this topic
and its interaction with other irradiation induced phenomena such as void swelling that necessitate
longer term study. A crucial uncertainty is how effective stress/strain relaxation is in a thick section
weld where the dose/flux dependence of irradiation creep varies significantly though the section
thickness. There is also a significant overlap with BWR interests where the long term values ofresidual stress are important for life assessments of core shroud structures.
[Scope]Just as with the proposed study of residual stress/strain profiles in thick section dissimilar metal
welds (PWR Outline Shorter Term Proposal 4), an alliance with a qualified fabricator of nuclear
components is essential in order to cover the correct ranges of welding methods, welding parameters,
weld profiles and weld constraint for austenitic stainless steel welds. Metallurgical characterization iscarried out by classical optical and electron microscopy and coupled with determination of residual
stress profiles by X-ray diffraction and hole drilling techniques, in this case before and after neutron
irradiation. Since only a few designs of PWR core support internals have welded structures close to
the nuclear fuel, the irradiation doses relevant to most welds in austenitic stainless steel are relatively
small in most PWRs but nevertheless in a range where relaxation by irradiation creep can besignificant. This study could be carried out in tandem with the one described in the preceding outline
proposal.
3.2.3 Initiation of SCC in Structural Alloys, Modeling Stochastic Features, Heat-to-Heat
Variability, Effects of Long Exposure Periods[Background]
There is an undoubted dearth of information relating to the effects of long term exposures andstochastic features of crack initiation in systems relevant to PWRs such as PWSCC of Alloy 600 and
the weld metals 132/182 and IGSCC of cold worked stainless steels. Such information is required tovalidate long term predictive models, identify any hitherto unanticipated environmentally induced
degradation phenomena, and bridge the gap to shorter term research work that is typically carried out
over periods of weeks or months. Another system requiring examination over the long term is IASCC
of reactor internals but this subject is dealt with under a separate heading related to neutron irradiatedaustenitic alloys.

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T. Shoji, et al./ Proactive Material Degradation Research Subjects for LWR

Evaluating SCC in structural alloys in order to quantify the effects of long term exposures andstochastic features of crack initiation automatically entails a significant long term commitment toexposure tests of many specimens under nominally identical test conditions for long periods of timemeasured in years. It also implies the duplication of significant numbers of test facilities and/or
development of test rigs capable of testing many specimens simultaneously. A parallel activity
requiring the cooperation of a suitably equipped hot laboratory capable of handling radioactively
contaminated artifacts and collaboration with a utility prepared to give access to components removed
from service would involve detailed examinations of oxide films and surfaces for signs ofenvironmentally induced deterioration or precursors likely to give rise to crack initiation such as grainboundary oxidation and grooving and pitting.
[Scope]The main systems considered important to study in the current context of very long term aging in
PWRs are PWSCC of Alloys 600, 132, 182 and 82 plus comparisons with Alloys 690, 152 and 52,
and IGSCC of cold worked austenitic stainless steels. This should include both the effects of bulk cold
work and surface cold work prior to tensile loading in the test facility and the effects of welding
defects such as hot cracks. While the mean behavior of any given SCC system under nominally thesame conditions can be determined with reasonable statistical confidence from less than 10 nominallyidentical specimens, evaluation of the standard deviation would require around 30 nominally identicalspecimens for each test condition.
Given present experimental capabilities identified in the published literature, it is probably feasible
to test around 30 small O-ring or C-ring specimens in chains in one experimental autoclave.
Monitoring displacements in such chains gives information on time to failure of individual specimens
although electrical continuity tests may also be feasible. Another technique based on testing multiple
tensile specimens where a small identity ball is displaced on failure of any given specimen is anotherpossible approach. However, tensile tests usually preclude the very high tensile strains/stresses thatcan develop locally in bending without plastic collapse and which are often the site of crack initiation.
The precise experimental strategy adopted would clearly require careful evaluation and prototype
testing.
In parallel, it is also proposed to examine any components removed from long term PWR service.
The aim would be to perform careful surface and cross-sectional examinations to search for signs oflong term deterioration at the oxide/metal interface likely to lead to crack initiation.

4. Research Subjects for Basic Research on Environmentally Assisted
Degradation of BWR and PWR Materials in common.
4.1 Research subjects for shorter term projects to be completed in ~5 years.
4.1.1 Effect of Environment on Fatigue and Fracture Resistance in BWRs and PWRs.
[Background]
The influence of LWR aqueous environments on fatigue endurance curves and fatigue crack
propagation rates for austenitic stainless steels and carbon and low alloy steels have received
considerable attention over the last three decades.

The main phenomenological responses toparameters such as cyclic frequency, strain rate during tensile loading, hold times, and corrosion
potential (i.e. as established in BWR NWC and HWC and PWR primary water conditions) have been50

T. Shoji, et al./ Proactive Material Degradation Research Subjects for LWR

well characterized for fatigue crack propagation. By contrast, little progress has been made in
fundamental understanding of the role of environment on fatigue crack initiation and growth,
especially in austenitic stainless steels and nickel base alloys, although somewhat better fundamental
understanding exists for carbon and low alloy steels where the role of sulfur based non-metallic
inclusions is known to predominate and affect localized dissolution/oxidation at cyclic strain
concentrations.
Two hypotheses invoking either anodic dissolution/oxidation at sites of cyclic strain concentrationor hydrogen effects for austenitic materials have been proposed and neither gives satisfactory
correlations for the effects of corrosion potential.

Consequently, empirical algorithms have been

imposed by regulatory authorities in Japan and the USA for austenitic stainless steels and nickel basealloys to take into account the effect of the environment on fatigue, particularly for the design process
to assess fatigue endurance. These algorithm have uncertainties in their long term application andparticularly in their application to complex strain histories associated with plant transients coupledwith realistic engineering surface finishes.
By contrast, the assessment of the environment on fracture resistance is a relatively new concern
where the basic phenomenology is still being developed for the complete range of LWR structural
materials. The topic can be divided into two parts; one is concerned with the apparent drop in JIc andtearing resistance during J-R tests of nickel base alloys at slow extension rates in PWR primary watercompared to conventional tests in air; the second concerns sudden ruptures of fracture mechanics test
specimens used for stress corrosion crack growth and corrosion fatigue crack growth testing in PWR
primary water at normal operating temperatures between 280 and 345°C and in BWR environments at
2880C.
In both cases, the strong presumption is that hydrogen absorbed from PWR primary water or BWR
HWC is responsible for reductions in fracture resistance. Note that even in oxygenated BWR NWC, asignificant fraction of the cathodic reaction is supported by hydrogen evolution (in addition to oxygen
reduction) that can be detected diffusing across the walls of thin capsules of stainless steel containing
high temperature oxygenated water. The hydrogen embrittlement hypothesis is reasonably well
supported by (a) results of J-R type tests conducted with different hydrogen concentration in the water,and (b), the correlation between increased susceptibility to reduced fracture resistance and thepresence of interfaces for hydrogen agglomeration such as semi-continuous carbide precipitates on
grain boundaries.
[Scope]The main requirement for further work in both of these areas of LWR environmental effects onmechanical properties of LWR structural materials, especially on fundamental mechanisms, comesfrom the need to determine the mechanism for the environmental effects on fatigue that will giveconfidence in the long term reliability of current assessment methods used during the design process,or modify them as appropriate. In the case of fatigue crack initiation and propagation in austenitic
stainless steels and nickel base alloys, there is a real need to establish the basic mechanisms ofenvironmental degradation of fatigue resistance. This is particularly the case in BWR HWC as well as
PWR primary and secondary conditions where the role of hydrogen concentration and corrosion
potential requires particular attention and characterization since these more reducing conditions are
known to be more detrimental for fatigue resistance than more oxidizing conditions encountered in51

T. Shoji, et al./ Proactive Material Degradation Research Subjects for LWR

BWR NWC. This concern is highlighted for BWRs when considering that many of these reactors are
adopting more reducing environments associated with "hydrogen water chemistry" and noble metal
technology.
In the case of fracture resistance, it is necessary to confirm the hydrogen embrittlement hypothesis,
to determine whether this is responsible for the sudden specimen ruptures at normal operating
temperatures mentioned above, and to determine a quantitative relationship between theconcentrations of hydrogen absorbed (since there will be gradients of hydrogen concentration in mostcomponents of practical interest), the relevant micro-structural features, and fracture resistance. The
proposition that the loss of fracture resistance during J-R testing is primarily due to environmentallyassisted cracking at the slow extension rates usually employed also needs to be confirmed since theobservation of sudden ruptures argues in favor of an intrinsic effect on mechanical properties due toexposure to the aqueous environment. As well as work on the nickel base alloys 600 and 690 and theirweld metals, there is also a need to determine to what extent the phenomenon extends to stainlesssteels, especially cast stainless steels (CASS) that may be thermally embrittled, and C&LAS.
4.1.2 Flow-Accelerated Corrosion in BWR and PWR Systems[Background]Flow-accelerated corrosion (FAC) describes enhanced corrosion that can occur in flowing water or
steam under single- or two-phase conditions.

It is associated with changes in the mass-transport

controlled oxidation and reduction electrochemical reaction rates occurring on an oxidized surface
next to the flowing water or steam.

FAC is most commonly observed in carbon steels used in thesecondary and tertiary circuits of PWRs and in the feedwater systems and steam lines of BWRs. It is
often associated with geometrical discontinuities or abrupt changes in flow direction where highturbulence is encountered. Typical metal losses from carbon steels can easily be >1mm/year for single
phase flow velocities >6 to 10 m/s at 200°C.
FAC can also occur on copper-base alloys used predominantly in condensers and for some tertiary
cooling water system components such as pumps. However, the use of copper-base alloys has been
widely phased out in favour of titanium alloys where seawater is used for cooling or stainless steels
where river or lake water cooling is used. Thus the scope of this section is confined to carbon steels.
The consequences of FAC include: (a) outright failure of the component due to thinning until thesystem pressure or stress can no longer be supported, (b) significant contribution to the feed wateriron content, (c) an increase in radiation levels in the balance-of-plant of BWRs and, (d) the fouling offlow measurement devices and ion exchange resins by released iron oxides. Extreme examples of theconsequences of FAC are the ruptures of an 18 inch diameter carbon steel condensate line on thesecondary side of the Surry-2 PWR in 1986 and a 24 inch diameter carbon steel steam line between
the low pressure heater and deaerator at the Mihama-3 PWR in 2004.
There are extensive data to indicate that the following system parameters, and their combinations, ,are of importance in determining the extent of FAC in carbon steels in water and steam;

temperature

oxygen and hydrogen concentration (corrosion potential)water chemistry and pHchemical composition of the substrate metal

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T. Shoji, et al./ Proactive Material Degradation Research Subjects for LWR

 single or two phase flow and turbulence.
This information has been sufficient to develop mitigation actions based on, for instance;(a)alloying additions (chromium >0.15-0.2wt.%); (b) adding oxygen >20-30ppb to BWR feedwaterlines, and (c) controlling the pH(25 o C) in PWR secondary systems close to 10.
In addition to developing mitigation actions, this phenomenological knowledge has been used for
decisions relating to inspection priorities.

Such an instance is the CHECWORKSTM prediction codedeveloped by EPRI. This code was developed after an extensive study of the factors controlling FAC
in LWRs. For each of the parameters affecting the phenomenology of FAC described above,
empirical predictive algorithms were developed. The code is widely used by LWR operators to predictthe most likely areas that may experience FAC, as well as the rate of metal loss in order to fix
inspection priorities. CHECWORKSTM is claimed to be accurate to within a factor of 2 whencompared with plant measurements, with a confidence level of 98%. A more recent evaluation of thiscode indicates, however, some discrepancies between observation and prediction.
An empirical approach to determining inspection priorities was also developed by NISA, following
the Mihama 3 accident. Thinning rates were examined by measuring the thicknesses of ~20,000
components at 23 PWRs and, with very few exceptions, the guidelines that had been developed as
early as 1990, were found to be conservative. BWR data were also examined and found to exhibit
lower corrosion rates than PWRs. This was attributed to water chemistry differences. Based on the
PWR inspections, the 1990 inspection rules were, nevertheless, revised with the inspection priorities
being governed by combinations of steam quality, velocity and temperature.
A similar empirical approach is taken in the COMSY code used in Germany, which is based on all
the known experimental and plant data on FAC. This code incorporates detailed modeling of the plantgeometry and thermal hydraulic characteristics and then, based on the water chemistry and material
compositions, evaluates the zones at risk from FAC.

For sub-systems identified as vulnerable to

FAC, a detailed analysis is performed to provide life predictions for individual components. An
integrated inspection management module enables inspection data to be incorporated as they become
available and optimize the inspection scope, locations and intervals.
The BRT-CICEROTM software developed by EDF is based on a physical model of ferrous ion
transfer between the boundary layer in equilibrium with magnetite reduction and the ferrous ion
concentration in the bulk water. The water chemistry and temperature are taken into account throughthe equilibrium ferrous ion concentrations in the boundary layer and that in the bulk water as well as
the thermal hydraulic parameters via the mass transfer coefficient.

Material composition parameters

affect primarily the oxide thickness and porosity in the model. The BRT-CICERO model was used
experimentally by EDF on a few plants in the early 1990’s but became mandatory for all EDF PWRs
after the discovery in 2001 of severe FAC on a reducer at the Fessenheim Unit 2 that had beencorrectly predicted by the model. Since then the model has been tested on over 6000 individual pipe
thickness measurements taken over the last 20 years on nearly 4800 different pipe elements in 58
plants. Recent improvements to the thermodynamics database, chemistry calculations (including
dissolved oxygen), alloy chemistry (specifically chromium), steam void calculation and manygeometrical factors for components have been updated. The agreement between observation (on morethan 14,000 measurements) and prediction are good; moreover where discrepancies do occur thepredicted metal loss is conservative in 93.3% of the comparisons.

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A less well known physical model was originally developed for the British Central Electricity
Generating Board to address FAC issues in the once through boilers of Advanced Gas Cooled reactors.
It has much the same features as the EDF model but, crucially, also takes into account the affect offlow on the cathodic reaction rate and the lowering of corrosion potential that ensues at high fluid
velocities.

This gives rise to a distinctly non-linear dependence of FAC kinetics at high masstransfer rates such that, in the specific AGR boiler problem addressed, the FAC rate depended on fluid[velocity]1.4. This extra effect of flow on the cathodic reaction seems to be an important effectrequiring much more attention.
[Scope]It is apparent from this background that caution must be given to any prediction capability that isbased solely on empirical data, without the support of a mechanistic base. This is especially the casefor situations, such as EAC and FAC where the inputs to the phenomenon involve many, interwoven
parameters. Such cautions are doubly important when attempting to extrapolate a prediction
capability beyond the extent of the empirical data base.
4.2 Research subjects for longer term projects to be completed in ~5 to 10 years.
4.2.1 Quantification of Potential Synergistic Effects between Competing Degradation Modes
for both PWRs and BWRs.
[Background]
Any study of material degradation mechanisms involving thermal or irradiation induced
embrittlement automatically involves a very long term commitment of time and resources. In addition,
long term thermal aging cannot be unduly accelerated by using high temperatures since there is a limitof about 400°C above which the physical aging mechanisms change.
The material where potential superposition of more than one aging process could be important iscast austeno-ferritic stainless steel (CASS) and stainless steel weld metals containing delta ferrite.
Large quantities of these materials are in use for the primary pipework, valve bodies and core support
structures beneath the core of PWRs and BWRs and welded structures. Several variables affecting thedegree of thermal aging of CASS and stainless steel weld metals have been extensively studied, suchas the percentage ferrite, chromium, molybdenum, and carbon contents, for their effect on fracture
toughness. However, very few studies have addressed the potential stress corrosion susceptibility ofthermally aged material in PWR or BWR primary water environments and even fewer the
combinations with radiation damage.
[Scope]The objective will be to study at least two heats of CASS with high susceptibility to thermal aging
and one with medium susceptibility. These materials would be thermally aged, probably at 400°C, forperiods up to those sufficient to simulate end of 40 and 60 operating lives. Note here that theactivation energy used for calculating equivalent thermal aging damage at lower temperatures may belower than commonly accepted, even as low as ~100 kJ/mole, and should be independently verified.
Irradiation up to 10 dpa could be done sequentially in a pool type thermal reactor and/or in a fast
reactor. It may also be possible to simultaneously thermally age at an elevated temperature up to
400°C and irradiate in a sodium cooled fast reactor with suitable shielding of the specimens to reduce
the fast flux to a value consistent with the likely doses in PWRs and BWRs (a few dpa) in the time54

T. Shoji, et al./ Proactive Material Degradation Research Subjects for LWR

required to simulate end-of-life thermal aging.
Following aging with various combinations of thermal and irradiation embrittlement, both fracture
toughness and resistance to stress corrosion cracking in simulated PWR and BWR primary coolants
would be evaluated.
4.2.2 Quantitative Modeling of Oxidation and EAC based on Fundamental PhysicalPrinciples for both PWRs and BWRs
[Background]
A very long term aim of research devoted to EAC of structural materials in light water reactor
systems is to develop predictive models based on fundamental atomistic physical principles. In thecontext of BWRs, the aim is to try and anticipate problems arising in primary circuit materials (andsteam generators) from very long term aging associated with plant lifetimes that are currently
anticipated to be prolonged from 40 to 60 years and eventually to 80 years and are clearly very
difficult to test experimentally without waiting for lead plants to attain such lifetimes.
The FRI, Tohoku University has already taken steps in the direction of more fundamental
physically based modeling through its work on the application of modern fracture mechanics solutions
for stress/strain fields at crack tips coupled with various hypotheses for crack advance, and in theapplication of molecular dynamics modeling to hydrogen and oxygen interactions with metals andoxygen. These initiatives respond in part to the desire to extend modeling to a priori atomistic
predictions of SCC susceptibility.
[Scope]
The following is a précis supplemented by points raised in discussion during the meeting of the
Tohoku-Hokkaido cluster held in Miyagi Zao during October 2010.
The surface enclave: smooth surfaces and imperfections; surface reactivity and films growth on
metals; misfit strains between oxides and underlying metal; atom removal stresses in films and in theunderlying metal.
The crack tip enclave: physical processes in tight nanometer scale wide cracks; vacancies, cavities,
dislocations and deformation at crack tips; crack tip chemistry; hydrogen activity and interactions;crack advance processes; short crack coalescence.
Long crack enclave: transition from slow to rapid (fracture mechanics controlled) crack growth; ion
migration, impurity hideout and crevice chemistry
Radiation enclave: environmental and metal interactions with neutrons and other ionizing radiations;
irradiation damage in metal and protective oxides; physical property changes; IASCC crack advanceprocesses.
SCC phenomenology enclave: This enclave is more component specific. For instance for PWRs theconcerns revolve around PWSCC (or LPSCC); Sulfur SCC, Lead SCC; SCC of low alloy steels andturbine alloys, effects of cold work and strain pathway. For the BWRs the concerns are primarily
related to IASCC of stainless steel core internal components, in addition to the balance of plant issues
identified for the PWRs.
The surface and radiation enclaves were judged in discussion at Miyagi Zao to be perhaps the least
well understood at this time.

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5. Conclusion
In this paper, research topics considered necessary for a proactive approach to aging degradation
phenomena in LWR structural materials which were discussed at the Proactive Aging Management
Experts' Panel Meeting of NISA Project, have been introduced. Several short term research subjects(to be completed or to achieve significant progress within 5 years) and several long term researchsubjects (to be completed in 5-10 years) were outlined for both PWRs and BWRs with a suggested
order of the priority. Although some may not necessarily be considered proactive in the sense thatthey have already been studied extensively by the industry and academia, they are included in thesuggested research subjects because of the great concern for the occurrence of the degradation incomponents that have not yet experienced it. It is also important to continue to promote the systematic
elicitation for potential and latent material aging degradation phenomena using the principles
developed in the Proactive Aging Management Experts' Panel Meeting.

6. References
1. Muscara, J “Expert Panel Report on Proactive Materials Degradation Assessment”, USNRCNUREG Report, CR-6923, February 2007.

ACKNOWLEDGEMENT
This work has been performed as part of the Program of Enhancement of Ageing Management andMaintenance of Nuclear Power Plants organized by the Nuclear and Industrial Safety Agency, NISA,
in the Ministry of Economy, Trade and Industry, METI, of Japan. The authors would like to thank theExperts' Panel members for their contributions without which no success of the program identifing,
discussing and prioritizing topics would never be made.
The Proactive Aging Management Experts' Panel Meeting Member in 2010
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Dr. Peter Ford (Consultant, GE CRD Retired, USA)
Dr. Roger W. Staehle (Consultant, USA)
Dr. Karen Gott (Consultant, SSM Retired, Sweden)
Dr. Raj Pathania (EPRI, USA)
Dr. Tiangan (TG) Lian (EPRI, USA)
Dr. Claude Amzallag (ONET – Technologies, France)
Dr. Jean-Paul Massoud (EdF SEPTEN, France)
Dr. Peter Scott (Consultant, AREVA Retired, France)
Dr. Peter Andresen (GE CRD, USA)
Prof. Hannu Hanninen (Helsinki U., Finland)
Dr. Armin Roth (AREVA, Germany)
Prof. Roger Newman (U. Toronto, Canada)Prof. Il Soon Hwang (SNU, the Republic of Korea)
Prof. En-Hou Han (IMR, China)
Dr. Dolores G. Briceno (CIEMAT, Spain)
Dr. Stephen M. Bruemmer (PNNL, USA)
Dr. G.E. (Gene) Carpenter (USNRC, USA)
Dr. Alan Turmbull (NPL, UK)
Prof. Robert Cottis (U. Manchester, UK)
Prof. Philippe Marcus (ENSCP, France)
Dr. Thierry Couvant (EdF Renardieres, France)
Dr. Torill M. Karlsen (OECD Halden, Norway)
Prof. Jean-Yves Cavaille (INSA-Lyon, France)
Prof. Yves Brechet (Institut Universitaire de France, France)
Dr. Pierre Combrade (Consultant, AREVA Retired, France)
Dr. Hans-Peter Seifert (PSI,Switzerland)

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Mr. M. Aoyama, Mr. Y. Sosa (NISA, Japan)
Prof. N. Sekimura (U.Tokyo, Japan)
Dr. S. Suzuki (TEPCO, Japan)
Mr. H. Tanaka (KEPCO, Japan)
Mr. T. Yaegashi (Tohoku-epco, Japan)
Mr. T. Kobayashi (MHI, Japan)
Dr. H. Fujimori (Hitachi-GE, Japan)
Dr. K. Arioka (INSS, Japan)
Dr. M. Akashi (Consultant, IHI Retired, Japan)
Associate Prof. T.Horiuchi (HIT, Japan)
Mr. H. Sakamoto (JNES, Japan)
Dr. K. Fukuya (INSS, Japan)
Mr. M. Ito (Toshiba, japan)
Dr. F. Inada (CREPRI, Japan)
Prof. Y. Ohki (Waseda U., Japan)
Prof. Y. Kitsutaka (Tokyo Metroporitan U., Japan)
Prof. H. Mihashi (Emritus P. Tohoku U, Japan)
Prof. T. Shoji (Tohoku U, Japan)
Prof. T. Kondo (Emeritus, Tohoku U, Japan)
Prof. Y. Watanabe (Tohoku U, Japan)
Associate Prof. Z. P. Lu (Tohoku U, Japan)
Associate Prof. Q. J. Peng (Tohoku U, Japan)
Associate Prof. M. Takahashi (Tohoku U, japan)
Assistant Prof. Y. Takeda (Tohoku U, Japan)
Specially Appointed Prof. T. Funada (Tohoku U, Japan)
Visiting Prof. J. Kuniya (Tohoku U, Japan)