Yves MEYZAUD, FRAMATOME-ANP (France)

15 Pages·2005·1.58 MB·English

by  Thibaut Lacheray

Y. MEYZAUD, A. LEFRANÇOIS, J.M. GRANDEMANGE,
FRAMATOME-ANP (France)
Aging of materials during plant operation
Preventive measures taken for EPR design
Y. Meyzaud, A. Lefrançois and J.M. Grandemange
Framatome-ANP, Tour AREVA, 92084 Paris La Défense, France
1. Introduction
The design of a new Pressurized Water Reactor (PWR) has to take into account past
experience, and in particular the lessons drawn from aging of components already in
operation. The benefit of this process for long term operation can be exceptionally high in the
case of an evolutionary design such as that of EPR, for which the materials and the
manufacturing conditions are the result of a very long term optimization.
Aging of materials in operation encompasses several damages, from the embrittlement caused
by irradiation or thermal aging to the generation of cracks by fatigue or corrosion. These
damages are known and characterized through international field experience and through a lot
of laboratory studies performed worldwide, in France mainly by EDF, CEA and Framatome-
ANP.
Even if some damage mechanisms are not fully understood, good practice rules or empirical
models were developed in order to give accurate predictions of damage kinetics and
consequences. All this experience was taken into account during the design phase of the EPR
project to define materials and/or service conditions allowing to prevent or to minimize aging.
This paper will present the rationale of the selection of materials which constitutes the first
step to prevent and minimize aging problems over a very long period. The materials have to
be well known to the manufacturers and have to present a good workability, inspectability and
weldability. The objective is here to manufacture high quality components with, as far as
possible, uniform microstructure and mechanical properties. It is also to avoid significant
fabrication defects which could obviously limit the lifetime of the components.
The second step of the process is the optimization of materials in order to prevent or at least to
minimize in service aging. For this, an in-depth understanding of the various damages
susceptible to occur in operation is necessary. Examples of the precautions taken for EPR
manufacturing will be given, regarding the effects of neutron irradiation, thermal aging,
fatigue and stress corrosion cracking.
At the end, it appears that EPR design benefits of the huge amount of R&D dedicated to aging
phenomena in PWRs, from which an optimization of materials and manufacturing is derived
to mitigate aging. It takes also benefit of the field experience to date and of the solutions
progressively developed to repair or replace damaged components. All these precautions
allow to be confident in the aging resistance of EPR materials for 60 years lifetime.
2. Principles of materials selection
(cid:1) General
Materials selection comes from the main features of PWRs, which can be summarized as
follows:
- the large size of many components or subassemblies such as the Reactor Pressure
Vessel (RPV), the reactor coolant pumps, the Steam Generators (SG), etc.
- the fabrication of these large components which entails forming, welding and cladding
operations. It is important to have easily weldable materials ensuring that the weld
metal and the heat-affected zone possess adequate properties.
- the condition of stainless steels and alloys which have to provide the best corrosion
resistance (no intergranular sensitization plus guaranteed surface cleanness) in
combination with very strict conditioning of the coolant to obviate stress corrosion
cracking.
- The regulatory requirements to prepare a "Materials File" for all pressure retaining
components in the main primary system. This requires a thorough knowledge of the
materials used and of the effects of the fabrication operations on their properties, so
that it can be demonstrated that they are suitable for application under all plant
operating conditions.
All this means that the materials selected are not necessarily "high performance" materials,
but rather commercial grades that are easy to use and well known to the manufacturers.
As a consequence, the requirements of the RCC-M and of most nuclear construction Codes
integrate:
- Narrower chemical analysis ranges for major components, reactor coolant piping and
steam generator tube materials.
- Very strict control of impurities and inclusions.
- Stringent non-destructive testing at all stages of manufacture.
- Detailed testing of the first fabricated component.
- Recording of essential variables governing materials properties in the supplier's
technical fabrication program.
(cid:1) Fast fracture
Among the regulatory requirements, one of the most important is the prevention of fast
fracture which can be defined as a fracture occurring without being preceded by any
significant overall deformation. As far as the materials are concerned, the resistance to fast
fracture is controlled through the measurement of:
- the Charpy V-notch toughness at several temperatures (at least 0°C) to check that the
ductile to brittle transition of ferritic steels is lower than the anticipated service
conditions and to avoid any catastrophic failure by cleavage.
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Aging of materials during plant operation - Preventive measures taken for EPR design
- The total elongation and reduction of area in tensile tests which can be correlated with
the resistance to ductile tearing.
In order to prevent fast fracture, stringent requirements apply to the materials at the
procurement stage. For example, for the main heavy components, additional margins towards
the fast fracture risk were introduced with a specified maximum RT of - 20°C. However
NDT
this is not enough, as some aging mechanisms are susceptible to reduce the resistance of
materials to fast fracture as a function of time, either by an increase of the ductile to brittle
transition temperature or by a decrease of the tearing resistance. These mechanisms, as well as
the measures which are taken to prevent or at least to minimize aging of materials, are
presented in paragraphs 3 to 6.
(cid:1) Corrosion
The corrosion behaviour of the materials is also an important factor not only from a plant
availability point of view but also from a corrosion products standpoint.
All surfaces in contact with primary water are clad with austenitic stainless steel or
constituted with austenitic stainless steel semi-finished products. The materials selected for
EPR have very low carbon contents or are stabilized with Titanium or Niobium, in order to
prevent any sensitization to intergranular corrosion during manufacturing.
Both austenitic stainless steel semi-finished products and weld overlay claddings must
guarantee a high level of protection not only against uniform surface corrosion, but also
against localized types of corrosion, such as stress corrosion cracking, intergranular corrosion,
pitting and crevice corrosion etc.. The undesirable effect of corrosion product transport is also
a relevant factor. Deposits of such products on fuel assemblies or on heat-exchanger surfaces
may lead to deterioration in heat transfer, or result in the formation of radioactive isotopes
after passing through the reactor core.
Concerning the limitation of the Cobalt content for reasons of dosimetry limitation, studies
were conducted in France and Germany, leading to optimized proposals considering the
release potential and the surfaces concerned. The surface consideration led to make a
particular effort for SG tubing, where a Co limitation to 0,015% is retained, the stainless
surfaces of the main components in contact with the fluid being limited to a Co content of
0,060%.
The release potential also led to a particular effort on the development of alternatives to
Cobalt-based hardfacings, in Germany and in France, which are proposed to the
manufacturers in particular for valves applications. Several Iron-based hardfacing alloys
(including "NOREM 02") and manufacturing processes were tested by Framatome-ANP and
EDF regarding the friction behaviour and the corrosion resistance in several environments
representative of normal operation and shutdown conditions. Design evolutions allowing the
suppression of the need for Cobalt-based hardfacings have also been evaluated.
There remain nevertheless some applications where Co-based hardfacings are necessary,
depending in particular on the function, the associated contact pressure and the temperature at
the time the function is required.
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Aging of materials during plant operation - Preventive measures taken for EPR design
(cid:1) Conclusion
The selection of materials for EPR derives from the selection previously made for the French
reactors N4 (tables 1 and 2) and for the German reactors Konvoi (tables 3 and 4). The
requirements for the EPR materials take into account the operating experience and the
evolution of the knowledge and as a consequence are more stringent than the previous ones.
3. Neutron Irradiation
Neutron irradiation of materials generates point defects caused by displacement in the crystal
lattice of atoms bombarded by incident neutrons. It also produces nuclear reactions leading to
the transmutation of some elements. The irradiation temperature is an essential parameter: at
low temperatures ((cid:1) 250°C) the potential for recombination of the point defects (and therefore
for "healing" of the material) is very limited; this possibility of thermal annealing increases
with temperature, so reducing the production of defects. At high temperatures (_ 400°C),
other phenomena may occur, like the formation of actual cavities or gas bubbles producing
swelling of the material. As the PWR service temperature is on the order of 300°C, the
behaviour of the materials is mainly influenced by the creation of point defects.
(cid:1) Reactor Vessel
In the RPV material, the residual elements, copper and phosphorus, interact with the point
defects to cause most of the brittleness. Apart from the chemical composition of the material
and its structure, the other significant parameters are the neutron dose and the temperature.
The flux and the neutron energy spectrum only seem to have second order effects.
The result for the RPV material submitted to a significant neutron fluence (core region) is an
increase in the yield strength, which raises the ductile-brittle transition temperature and thus
entails an increased risk of fast fracture. The factors determining the amount of irradiation
embrittlement are well known. For the RPV materials, these factors are the concentrations in
residual elements, copper and phosphorus. Some alloying elements (Ni, etc.) may reinforce
the effect of the residual elements. The residual elements causing most of the embrittlement
have long been strictly limited by the specifications. In addition, during the reactor vessel
design stage, an analysis of the risk of fast fracture of the reactor vessel is performed using
formulas predicting irradiation embrittlement. The actual embrittlement of the materials is
measured periodically by mechanical tests carried out on test specimens inserted inside the
reactor vessel and subjected to a greater neutron flux than the vessel wall.
As for EPR, due to the large diameter of the reactor vessel, the neutron fluence on the wall is
reduced to very low values, like in the case of German Konvoi plants (1.25 to 2.5 x 1019
neutrons/cm2, E > 1 MeV, .for 60 years, depending on fuel loading conditions). The
requirements relative to the maximum contents in phosphorus and copper are reinforced. The
final result is an anticipated ductile to brittle transition temperature RTNDT of less than 30 °C
after 60 years service.
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Aging of materials during plant operation - Preventive measures taken for EPR design
(cid:1) RPV Internals
The internal structures surrounding PWR cores are made from austenitic stainless steel. They
are cooled by the primary water, but (cid:1) heating means that their operating temperature is
between 300 and 400°C. Very high neutron doses are received and in certain regions these
may approach 1 x 1023 neutrons/cm2 (with E > 1 MeV). The structure of austenitic stainless
steels makes them much more resistant to irradiation than the vessel steel. However, the
neutron doses experienced by the internal structures are so high that the properties of these
materials may undergo profound changes. The first visible effect is a hardening that increases
with falling irradiation temperature. Important modifications to the microstructure,
particularly near the grain boundaries, are caused by diffusion of the alloying elements under
irradiation and this may make the material susceptible to various forms of corrosion.
Figure 1: Schematic of lower EPR internals (heavy reflector)
Thus intergranular corrosion has been observed in the internals of boiling water reactors, even
for relatively moderate neutron doses. IGSCC has also been associated to local rough grinding
or heavy cold work. For PWRs, the lack of oxygen in the primary water obviates this risk.
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Aging of materials during plant operation - Preventive measures taken for EPR design
The only degradation so far observed is the rupture of some heavily irradiated internal
structures fastening bolts located in enclosed spaces in old plants.
International research programs are underway to improve our knowledge of the behaviour of
materials used at present and to identify the most resistant materials, which could be
employed for repairs or replacements.
For EPR, austenitic stainless steel heavy reflectors are selected for the RPV lower internals,
which reduce fast neutron leaks and contribute to the reduction of the fluence on the RPV
wall. The design uses forged blocks which are held with tie-rods away from the core in order
to reduce the risk of having welds or bolting submitted to high neutron fluence.
4. Thermal aging
In some materials a diffusion-controlled precipitation mechanism is active even at service
temperatures. This process called thermal aging leads to a loss of ductility, deformability and
toughness. The selection of material is limited to materials that are not or not too much
susceptible to this effect at service temperature.
The conditions required for thermal aging include having an "unstable" material (with a
quenched structure in particular) in which the atoms are able to rearrange themselves by
diffusion. The rate of diffusion is higher for bainitic low-alloy steels or ferritic and martensitic
(body-centered cubic) stainless steels than for austenitic steels or alloys (face-centered cubic
structure). Steels in the first group are therefore more affected by aging. The two main
phenomena causing thermal aging are: (a) intergranular segregation of phosphorus in
martensitic and bainitic steels and (b) "unmixing" of chromium from its solid solution in the
ferrite of duplex austenitic-ferritic stainless steels and in martensitic stainless steels.
Austenitic stainless steels can be assumed to be unaffected by thermal aging. In the special
case of type 600 or 690 nickel based alloys, comprehensive work has been done by
Framatome-ANP to show that aging due to long-range ordering does not occur in commercial
materials.
(cid:1) Low alloy steels
The low-alloy Mn-Ni-Mo steels used to make the reactor pressure vessels, steam generators
and pressurizers are susceptible to intergranular embrittlement by segregation of phosphorus
and other impurities (Sn, Sb and As). For the base material or deposited weld metal, the shift
of the transition temperature for the Charpy V-notch impact test is not more than 30°C. On
the other hand, considerable embrittlement has been observed for higher service temperatures
(325 to 350°C) in the large-grained areas of zones (HAZs) affected thermally by welding
operations. This risk is minimized by the measures that have long been applied in fabrication
to avoid having large-grained HAZs subsisting locally in the case of joining or cladding
operations on large components.
Since the 1970s, owing to the progress made in steelmaking, reactor vessel steels became
cleaner, notably in terms of reduced sulphur and phosphorus contents. The consistency of
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product analysis has also improved considerably, and today the result is excellent
reproducibility of fabricated items. The 16 MND 5 steel defined in the 2000 edition of the
RCC-M is very close to its optimum level. For the EPR application, the phosphorus content is
reduced further by limiting this element to 0.008 %, thus reducing the material's sensitivity to
thermal aging.
(cid:1) Cast duplex austenitic-ferritic stainless steels
The cast duplex austenitic-ferritic stainless steels are used to make primary loop piping
(centrifugally-cast straight sections and cast elbows) and pump casings. The grades used
(RCC-M Z 3 CN 20-09 M, type CF8, and Z 3 CND 19-10 M, type CF8M) are derived from
the standard austenitic stainless steels (type 18-10 and 17-12 Mo respectively). The cast steels
have a dual phase (duplex) structure consisting of (cid:1) ferrite (typically 10 to 25 %) in an
austenitic matrix. The ferrite is there to forestall any risk of hot cracking during casting and to
raise the tensile properties up to the level of forged or rolled (wrought) austenitic steels. It
also improves the resistance to different forms of corrosion. The ferrite phase can be
considered to be continuous; this has been shown in particular by the experiments to
selectively dissolve austenite.
During aging, a spinodal decomposition of the chromium-iron solid solution hardens the
ferrite (its hardness may reach in the worst cases a Vickers value of 600 to 800) and makes it
susceptible to cleavage fracture, even at a temperature as high as 300°C.
The following important embrittlement parameters have been found from the research carried
out:
- Aging temperature and time.
- Chemical composition and ferrite content: High Cr, Si, Mo and ferrite contents are
unfavourable.
Using the available data, it has been possible to establish embrittlement prediction formulas
employing the chemical composition, the ferrite content and the operating temperature and
duration.
Highly variable aging sensitivities are encountered within the ranges laid down in casting
grade specifications, depending on the chemical composition and service temperatures. It can
therefore be seen that the consequences of aging in service cannot be considered
comprehensively: each component must be treated as a special case depending on the
fabrication and service conditions.
The toughness of aged duplex austenitic-ferritic steels can be characterized using parameters
derived from nonlinear fracture mechanics, J or J and dJ/da. The lower limit toughness
IC 02
values for aged austenitic-ferritic stainless steels have been determined. The validity of the
predictions based on laboratory work is being confirmed by appraisals of components taken
out of service. This work was started on some elbows removed when steam generators were
replaced.
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Aging of materials during plant operation - Preventive measures taken for EPR design
Compared with the materials of the 1970s, austenitic stainless steels have above all improved
in cleanliness from inclusions (the sulphur content has been cut in practice from 0.015 to
0.003 %). The reproducibility of product analysis, owing to advances in steelmaking has also
been improved markedly. As far as cast austenitic-ferritic steels are concerned, consideration
of in-service thermal aging led to dropping the Z 4 CND 19-10 M grade (CF8M type) at the
beginning of the 1980s for applications above 250°C, since it was most sensitive to aging. For
similar reasons, the ferrite content of the Z 4 CN 20-09 M grade (CF8 grade) was limited to
ensure good end-of-life toughness. For primary loop elbows, the final development has been
the use of bent forged steel pipes for Civaux N4 plant.
It can therefore be considered that the austenitic stainless steels grades, as well as the
molybdenum free Z4 CN 20-09M duplex austenitic-ferritic casting grade, have been
completely optimized. In the case of EPR, the application of the RCC-M Z4 CN 20-09M
grade is restricted to the primary pump casing which operates at the cold leg temperature. A
huge amount of data relative to thermal aging of this grade is available in France. In
particular, long term thermal aging up to 100 000 h at 300 and 350°C, which represents more
than 100 years operation at the cold leg temperature, has been applied to numerous pump
casing samples to determine the lower bound properties of the Z4 CN 20-09M grade. As an
example, in the worst aging conditions, the minimum tearing resistance parameter J is of the
0.2
order of 200 kJ/m2 at the operating temperature, which can be considered as satisfactory,
The EPR primary piping loops will be manufactured with forged austenitic stainless steels, a
solution which suppresses totally thermal aging susceptibility.
Figure 2: Example of forged cold leg with integral elbow and nozzle
(cid:1) Martensitic stainless steels
Forged martensitic stainless steels are used in PWRs mainly for internal and external bolting
(bolts, studs and nuts) and to make valve stems and some parts of control rod drive
mechanisms. The main grades contain either 13 % Cr (RCC-M Z 12 C 13 and Z 12 CN 13)
or 16 % Cr (Z 5 CND 16-4 and Z 5 CNU 17-4). Precipitation of the chromium-rich (cid:1)' phase
is responsible for the hardening of these alloys during the aging process. For structural-
hardening steels, like 17-4 PH (Z 5 CNU 17-4), aging leads to additional precipitation of the
hardening phase (here (cid:2)) within short time periods and then coalescence of all the precipitates
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of the same phase. For steels not containing molybdenum (Z 12 C 13, Z 12 CN 13 and Z 5
CNU 17-4), an intergranular embrittlement phenomenon due to phosphorus segregation may
occur in isolation or be superimposed on the other embrittlement mechanisms.
These materials age relatively quickly. Two years at 350°C are enough to reach maximum
embrittlement. To a first approximation, the hardening or the transition temperature shift
caused by (cid:1)' precipitation follow a time-temperature equivalence relationship.
The effects of aging were revealed when several valve stems ruptured in difference places in
the world when operated cold.
As a consequence, martensitic stainless steels containing more than 13 % Cr have to be
avoided for permanent operation above 250 °C, if submitted to significant stresses at low
temperatures. In order to be able to replace the 17-4 PH grade in some applications, a
cooperative study was performed by EDF and Framatome-ANP to develop a replacement
grade which remains ductile at room temperature even in the fully aged condition. This grade
will be introduced in the next edition of the RCC-M Code.
5. Fatigue
Another well known mechanism for crack initiation is the fatigue damage.
Cyclic loading can influence the lifetime of components. This is due to the fact that
alternating stresses and the resulting strains affect the microstructure of metallic materials.
The accumulated fatigue strain leads to re-arrangement of the dislocation network and
increase of the density of dislocations in the material microstructure, formation of slip bands
at the surface and then of micro cracks.
This mechanism appears in two steps: firstly an initiation of small cracks in the most stressed
zones, and then a crack propagation stage which may lead to a leak or to a fast fracture risk if
the applied loading is sufficient. This phenomenon is essentially related to the mechanical and
thermal loads, the environment affecting the material resistance to crack initiation and
propagation. These initiation and propagation phases may be observed on sound structures in
highly stressed zones or in a structure with a pre-existent defect.
The prevention of fatigue crack initiation involves a certain number of state of the art rules for
the design of circuits and the drawing of parts, in particular in transition radii, and by doing a
detailed fatigue analysis in the most stressed zones, taking into account all planned in-service
conditions. In-service load follow ensures that the loading effectively applied is enveloped by
the design hypotheses and that the validity of the integrity demonstration may be ensured for
the entire service life of the equipment. Where appropriate, specific surface roughness control
may be imposed by the specification to further extend the fatigue resistance.
As a result of these precautions, fatigue problems do not generally concern zones subjected to
large stress cycles, subject to detailed analyses and in-service surveillance. They concern
fatigue damages under high cycle fatigue conditions (vibrations or local thermal-hydraulic
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Aging of materials during plant operation - Preventive measures taken for EPR design
phenomena), which were not identified during the pre-operational tests or which result from
malfunctions:
- Thermal fatigue cracking of the ferritic base metal close to the feedwater pipe
connection to the steam generator nozzle, In this location, the environment can reduce
the resistance of the material to low cycle fatigue, except when a careful monitoring of
the quality of the feedwater is performed. It is the case for EPR where, in addition, the
sulphur content of the feedwater line is kept below 0.005 % to prevent the occurrence
of any corrosion fatigue
- Vibration of instrumentation piping not sufficiently supported, as a result of
instabilities in fluid flow, and possible mechanical amplification by mechanical
resonance function of the circuit configuration: these problems are solved by
installation recommendations and more attention paid to pre-operational test results.
- Thermal fatigue in mixing zones between hot and cold water: examples are thermal
barriers in primary pumps and auxiliary mixing zones.
- Stratification or vortex in dead legs of circuits, between two isolation valves or
between the main pipe and an isolation valve on a branch pipe. Such problems are
solved by improving the circuit design: separation of normal and emergency feedwater
systems on steam generators, minimum slope imposed on the surge line sections,
venting or pressurization of pipe portions situated between two isolation devices,
suppression of cold leaks coming from non leaktight isolation valves etc.
These lessons from the field are taken into account for the design of EPR. For detailed fatigue
analyses, considering the good operating experience of components for which fatigue
evaluations were made at the design stage, the methodology of calculation of the usage factor
is considered conservative and kept unchanged.
6. Stress Corrosion Cracking
Stress corrosion cracking may appear in service according to one of the following
mechanisms, under high service or residual manufacturing stresses:
- Strain-induced Corrosion Cracking (SICC) of carbon or low alloy steels can occur if
the following criteria are met:
• strain rates in a range of 10-5 to 10-7 sec-1 with localized yielding,
• high oxygen content in the water phase > 50 ppb under stagnant conditions (in
case of flowing condition the threshold value is higher),
• temperature range of approximately 150°C to 300°C with a maximum sensitivity
near by 240°C.
- Stress corrosion by chlorides, which provoke transgranular cracking of austenitic
stainless steels. Where these steels are sensitized, the cracking may also be
intergranular. This form of corrosion is likely to develop in environments containing
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Aging of materials during plant operation - Preventive measures taken for EPR design