In this paper, nuclear power plant
concrete structures are described and
their operating experience noted. Primary
considerations related to management
of their aging are noted: degradation
mechanisms, damage modeling,
and material performance; assessment
and repair; and applications of structural
reliability theory. Several topics
are noted where additional research
would be of benefit.
INTRODUCTION
In the United States, the Atomic Energy
Act and regulations of the United
States Nuclear Regulatory Commission
(USNRC) limit commercial power
reactor licenses to an initial 40-year period,
but also permit such licenses to be
renewed. Other countries may not have
a limit set on the plant operating license
period but the utility must obtain a permanent
renewal of its operating license
subject to numerous and continuous
justifications (e.g., periodic safety reevaluations).
This 40-year term for reactor
licenses was based on economic
and antitrust considerations—not on
limitations of nuclear technology. Due
to this selected period, however, some
structures and components may have
been engineered on the basis of an expected
40-year service life.
Currently the United States has 104
nuclear power plant units licensed for
commercial operation that provide
about 20% of the electricity supply.
The first of these units is nearing the
end of its initial operating license period. In order to help assure an adequate
energy supply, the USNRC has established
a timely license renewal process
and clear requirements that are needed
to ensure safe plant operation for an
extended plant life. These requirements
are codified in Parts 51 and 54 of Title
10, “Energy,” of the Code of Federal
Regulations (CFR) and provide for a renewal
of an operating license for an additional
20 years. In order to ensure the
safe operation of nuclear power plants,
it is essential that the effects of age-related
degradation of plant structures,
as well as systems and components, be
assessed and managed during both the
current operating license period as well
as subsequent license renewal periods.
HOW WOULD YOU... |
…describe the overall significance
of this paper?
This paper provides a description
of nuclear power plant concrete
structures, their operating
experience, potential degradation
mechanisms, approaches
for assuring their continued
performance, and areas where
additional research would be
beneficial.
…describe this work to a
materials science and engineering
professional with no experience in
your technical specialty?
This paper describes large concrete
structures that provide safety
functions in nuclear power plants,
their performance, potential factors
that could infl uence their continued
reliable performance, methods for
estimating future performance, and
desired research to provide added
assurance of continued reliable and
safe service.
…describe this work to a
layperson?
Concrete structures in nuclear
power plants are described, a review
of their operational experience
provided, potential environmental
effects that could potentially impact
their performance provided, and
methods described that can identify
and mitigate any environmental
effects that could impact continued
reliable and safe performance of the
concrete structures. |
CONCRETE STRUCTURES AND OPERATING EXPERIENCE
Concrete Structures
All commercial nuclear power plants
in the United States contain concrete
structures whose performance and
function are necessary for protection
of the safety of plant operating personnel
and the general public, as well as
the environment. The basic laws that
regulate the design (and construction)
of nuclear power plants are contained
in Title 10 of the CFR, which is clarified by documents such as Regulatory
Guides, U.S. Nuclear Regulatory Commission
NUREG reports, and Standard
Review Plans.
A myriad of concrete-based structures
are contained as part of a lightwater
reactor (LWR) plant to provide
foundation, support, shielding, and
containment functions. Typical safety-related
concrete structures contained in
LWR plants may be grouped into four
general categories: primary containments,
containment internal structures,
secondary containments/reactor buildings,
and other structures. Only information
related to primary containment
structures for pressurized-water (PWR)
and boiling-water reactor (BWR) plants
is summarized here. (Information on
other concrete structures is provided
elsewhere.1)
Of the PWR plants that have been licensed
for commercial operation in the United States, approximately 80% use
either reinforced or prestressed concrete
primary containments. The concrete
containments are of three different
functional designs: subatmospheric
(reinforced concrete), ice condenser
(reinforced concrete), and large/dry
(reinforced and prestressed concrete).
The primary differences between these
containment designs relate to volume
requirements, provisions for accident
loadings/pressures, and containment
internal structures layout. The PWR
containment structure generally consists
of a concrete basemat foundation,
vertical cylindrical walls, and dome.
Leak tightness of a containment is provided
by a steel liner attached to the
containment inside surfaces. Exposed
surfaces of the carbon steel liner are
typically painted to protect against corrosion
and to facilitate decontamination
should it be required. Depending
on the functional design (e.g., large dry
or ice condenser), the concrete containments
can be on the order of 40 m to
50 m in diameter and 60 m to 70 m
high, with wall and dome thicknesses
from 0.9 m to 1.4 m, and base slab
thicknesses from 2.7 m to 4.1 m. Figure
1 presents the Trojan nuclear plant
cooling tower and post-tensioned concrete
containment prior to decommissioning
and demolition. Pressurized-water
reactor plants that utilize a metallic
primary containment (large dry
and ice condenser designs) are usually
contained in reinforced concrete “enclosure”
or “shield” buildings that, in
addition to withstanding environmental
effects, provide radiation shielding and
particulate collection, and ensure that
the free-standing metallic primary containment
is protected from the natural
environment.
Of the BWR plants in the United
States, approximately 30% utilize either
reinforced or prestressed concrete
primary containments. Boiling-water
reactor containments, because of
provisions for pressure suppression,
typically have “normally dry” sections
(dry well) and “flooded” sections (wet
well) that are interconnected via piping
or vents. Boiling-water reactor plants
that utilize steel primary containments
have reinforced concrete structures that
serve as secondary containments or reactor
buildings. These structures generally
are safety-related because they
provide additional radiation shielding;
provide resistance to environmental and
operational loadings; and house safetyrelated
mechanical equipment, spent
fuel, and the primary metal containment.
Although these structures may
be massive in cross section in order to
meet shielding or load-bearing requirements,
they generally have smaller elemental
thicknesses than primary containments
because of reduced exposure
under postulated accident loadings.
Operating Experience
In general, the performance of nuclear
power plant safety-related concrete
structures has been very good.
However, there have been a few isolated
incidences of degradation that primarily
occurred early in life and have
been corrected.2 Causes generally were
related either to improper material selection
or construction/design deficiencies.
Examples of some of these problems
include low concrete compressive
strengths, voids under the post-tensioning
tendon bearing plates, cracking of post-tensioning anchor heads, containment
dome delaminations, misplaced
steel reinforcement, post-tensioning
system button-head deficiencies, and
water-contaminated corrosion inhibitors.
Several incidences of degradation
related to environmental effects have
occurred. Examples include corrosion
of steel reinforcement in water intake
structures, corrosion of post-tensioning
tendon wires, leaching of tendon gallery
concrete, low prestressing forces,
and leakage of corrosion inhibitors
from tendon sheaths. Other aging-related
problems include cracking and
spalling of containment dome concrete
due to freezing and thawing, and corrosion
of containment liners. As the
plants age incidences of degradation
are expected to increase, primarily due
to environmental effects. Additional
information on degradation of U.S.
nuclear power plant concrete structures
is available,3,4 as well as problem areas
experienced with nuclear power plant
concrete structures in other countries.5
AGING MANAGEMENT
Guidelines for production of durable
concrete are available in national consensus
codes and standards (e.g., ACI
318)6 that have been developed over the
years through knowledge acquired in
testing laboratories and supplemented
by field experience. Serviceability of
concrete has been incorporated into the
codes through strength requirements
and limitations on service load conditions
in the structure (e.g., allowable
crack widths, limitations on mid-span
deflections of beams, and maximum
service level stresses in prestressed
members). Durability has been included
in the design through specifications
for maximum water-cement ratios, requirements
for entrained air, minimum
concrete cover over reinforcement, etc.
Service-related degradation, however,
can affect the performance of nuclear
power plant concrete structures. As
these plants mature, environmental factors
are going to become increasingly
important. Demonstration of continued
safe and reliable operation of the plants
will involve implementation of a program
that effectively manages aging to
ensure the availability of design safety
functions throughout the plant service
life.
General guidance on developing an
aging management program for concrete
containment buildings has been
developed.5 Additional information is
available through organizations such as
The Electric Power Research Institute,
The International Union of Laboratories
and Experts in Construction Materials,
Systems and Structures, and the
Nuclear Energy Agency Committee
on the Safety of Nuclear Installations
under its Integrity of Components and
Structures Working Group.
DEGRADATION MECHANISMS, DAMAGE MODELING, AND MATERIAL PERFORMANCE
Degradation Mechanisms
Exposure to the environment (e.g.,
temperature, moisture, cyclic loadings,
etc.) can produce degradation of reinforced
concrete structures. The rate of
deterioration is dependent on the component’s
structural design, materials
selection, and quality of construction,
curing, and aggressiveness of environmental
exposure. Termination of a
component’s service life occurs when
it no longer can meet its functional and
performance requirements.
Primary mechanisms (factors) that,
under unfavorable conditions can produce
premature deterioration of reinforced
concrete structures include
those that impact either the concrete
or steel reinforcing materials (i.e., mild
steel reinforcement or post-tensioning
systems). Degradation of the concrete
can be caused by adverse performance
of either its cement-paste matrix or aggregate
materials under chemical or
physical attack. In nearly all chemical
and physical processes influencing
the durability of concrete structures,
dominant factors involved include the
transport mechanisms within the pores
and cracks, and the presence of water.
Degradation of mild steel reinforcing
materials can occur as a result of corrosion,
irradiation, elevated temperature,
or fatigue effects, with corrosion being
the most likely form of attack. Posttensioning
systems are susceptible to
the same degradation mechanisms as
mild steel reinforcement plus loss of
prestressing force, primarily due to tendon
relaxation, and concrete creep and
shrinkage. Additional information on
durability of nuclear power plant reinforced
concrete structures is available.7
Damage Modeling
Extensive research and studies have
been carried out to determine the durability
of concrete under various service
conditions, and thus information on
the progressive changes in the physical
and chemical nature of concrete under
such conditions is available.8 Damage
models for reinforced concrete in large
measure have addressed corrosion of
steel reinforcement and new construction.
Improved damage models and
guidelines for their use are required to
predict failure probability of a degraded
concrete structure, either at present
or at some future point in time. Synergistic
effects involving more than one
degradation factor and the interaction of loading and environment also need
to be investigated in more detail.
Material Performance
Nuclear safety-related concrete
structures are composed of several
constituents that, in concert, perform
multiple functions (e.g., load-carrying
capacity, radiation shielding, and
leak tightness). Primarily, these constituents
include the following material
systems: concrete, conventional steel
reinforcement, prestressing steel, steel
liner plate, and embedment steel. Data
on the long-term performance of the
reinforced concrete materials is of importance
for demonstrating the durability
of the nuclear power plant concrete
structures and in predicting their performance
under the influence of pertinent
aging factors and environmental
stressors. This information also has application
to establishing limits on hostile
environmental exposure for these
structures and to the development of
inspection and maintenance programs
that will prolong component service
life and improve the probability of the
component surviving an extreme event
such as a loss-of-coolant accident.
Reviews of research conducted on
concrete materials and structures indicate
that only limited data are available
on the long-term (40 to 80 years)
properties of reinforced concrete materials.2 Where concrete properties have
been reported for conditions that have
been well documented, the results were
generally for concretes having ages < 5
years, or for specimens that had been
subjected to extreme, nonrepresentative
environmental conditions such as
seawater exposure or accelerated aging.
Relatively few investigations were
reported providing results on examinations
of structures that had been in
service for the time period of interest,
20 to 100 years, and they did not generally
provide the “high quality” baseline
information (e.g., baseline material
characteristics and changes in material
properties with time) that is desired
for meaningful assessments to indicate
how the structures have changed under
the influence of aging factors and environmental
stressors.
Limited data on the long-term performance
of reinforced concrete materials
reported in the literature, results from
concrete cores removed from nuclear
power plants, and specimens cast in
conjunction with nuclear power plant
facilities have been reported.9 As noted
in Figure 2, these results generally show
an increase in compressive strength
(relative to 28-d reference strength) at
a decreasing rate with age, but the data
obtained from the literature were for
concrete ages ≤ 50 years and the nuclear
plant data for ages ≤ 27 years. With the
availability of decommissioned nuclear
power plants and plant modifications
requiring removal of materials, opportunities
exist to obtain samples for use
in providing an improved understanding
of the effects of extended exposure
under the conditions found in nuclear
power plants. Results could then be input
into an operational database to help
monitor and benchmark specific plant
performance. Examples of areas of interest
would be the effects of long-term
thermal loadings at moderate temperature
levels and effects of irradiation on
load-bearing concrete structures operating
> 40 years. Additional applications
of a concrete material sampling
activity would be for assessment of
construction quality, development of
improved damage models, assessment
and validation of non-destructive testing
methods, and evaluation of the performance
of repair activities.
With respect to ungrouted post-tensioning
systems, current examination
programs are adequate for determining
the condition of the post-tensioning
system materials and evaluating the
effects of conventional degradation.
Isolated incidences of wire failure due
to corrosion have occurred. Leakage
of tendon sheathing filler has occurred
at a few plants, but except for the potential
loss of corrosion protection, the
problem appears to be primarily aesthetic.10 Tendon forces at most plants
are acceptable by a significant margin,
but larger than anticipated loss of force
has occurred at a few older plants. Although
bonded prestressing tendons
are less vulnerable to local damage
than ungrouted tendons, the grouted
tendons can not be visually inspected,
mechanically tested, or re-tensioned in
the event of larger than anticipated loss
of prestress. Improved guidance on inservice
inspection of grouted tendons is
desired. Other potential research topics
related to post-tensioning systems include
development of an improved relationship between the end-anchorage
force measured by the lift-off test and
change in mean force along the tendon
length for unbonded tendons.
ASSESSMENT AND REPAIR
Operating experience has demonstrated
that periodic inspection, maintenance,
and repair are essential elements
of an overall program to maintain an
acceptable level of reliability for structures
over their service life. Assessment
and management of aging in
nuclear power plant concrete structures
requires a more systematic approach
than simple reliance on existing code
margins of safety.11 What is required
is the integration of structural component
function, potential degradation
mechanisms, and appropriate control
programs into a quantitative evaluation
procedure. A methodology to demonstrate
the reliable and safe performance
of these structures should include
identification of structures important
to public health and safety; identification
of environmental stressors, aging
mechanisms and their significance, and
likely sites for occurrence; a monitoring
or in-service-inspection-based
methodology that includes criteria for
resolution of existing conditions; and a
remedial measures program.
Component Selection
The most effective structural condition
assessment programs are those
that focus on the components most
important to safety and at risk due to
environmental stressor effects. Aging
assessment methodologies have been
developed to provide a logical basis for
identifying the critical concrete structural
elements and degradation factors
that can potentially impact the performance
of these structures.12,13 An evaluation
of the impact on plant risk due to
structural aging can also be used in the
selection of structural components for
evaluation.14 A combination of finiteelement
analysis and nondestructive
testing methods for evaluation of aging
and degradation of concrete containments
has also been proposed as a basis
for component selection.15
In-service Inspections
In-service inspection programs are
conducted to help ensure that the nuclear
power plant structures have sufficient
structural margins to continue to
perform in a reliable and safe manner.
A secondary goal is to identify environmental
stressors or aging factor effects
before they reach sufficient intensity
to potentially degrade structural components.
Routine observation, general
visual inspections, leakage-rate tests,
and destructive and nondestructive examinations
are techniques available to
identify areas of nuclear power plants
that have experienced degradation.
Determining the existing performance
characteristics and extent and causes of
any observed distress is accomplished
through a structural condition assessment
that routinely initiates with a
general visual inspection to identify
suspect areas followed by application
of destructive or nondestructive examinations
to quantify the extent and significance
of any observed degradation.
More detailed information on guidelines
on conduct of surveys of existing
civil engineering buildings is available.16 Figure 3 presents an example of
a survey to locate and characterize steel
reinforcement embedded in a general
civil engineering concrete structure.
Some general guidance on assessment
of nuclear power plant degradation
is also available.17 However, nuclear
power plant reinforced concrete
structures present special challenges
for development of acceptance criteria
because of their massive size, limited
accessibility in certain areas, stochastic
nature of past and future loads,
randomness in strength, uncertainty in
material changes due to aging and possibly
degradation, and qualitative nature
of many nondestructive evaluation
techniques. Improved guidelines and
criteria to aid in the interpretation of
condition assessment results, including
development of probability-based degradation
acceptance limits, are desired.
Nondestructive Examinations
Application of nondestructive examination
methods to nuclear power plant
reinforced concrete structures presents
challenges: wall thicknesses can be
in excess of one meter; structures often
have increased steel reinforcement
density with complex detailing; there
can be a number of penetrations or
cast-in-place items; accessibility may
be limited due to the presence of liners
or other components, harsh environments,
or structures may be located
below ground; experience with nondestructive
examinations of nuclear power
plant concrete structures is somewhat
limited; and methods utilized
for the nuclear power plant structures
are often based on equipment developed
for other materials or technologies.
Available methods are relatively
good at identifying cracking, voids,
and delaminations, and indicating the
relative quality of concrete. Methods
for determining concrete properties,
however, are somewhat more qualitative
than quantitative because they tend
to be indirect in that they often require
the development of correlation curves
for relating a measured parameter (e.g.,
ultrasonic velocity or rebound number)
to a property (e.g., concrete compressive
strength). Identification and description
of methods for determining
the strength of concrete and evaluation
of concrete structures is provided elsewhere.18,19 A practical guide related to
nondestructive examination of concrete
has been developed that not only identifies
and describes the capabilities, limitations,
and applications of the various
methods that are available, but presents
results from a number of examples.20
Noninvasive techniques for characterization,
inspection, and monitoring
of thick-walled, heavily reinforcedconcrete
structures in nuclear power
plants to provide additional assurances
of their continued structural integrity
are desirable (e.g., as-built or current
structural features determination, flaw
detection and characterization, identification
of honeycomb areas and embedded
items, and location of voids
adjacent to the liner). Methods that can
be used to inspect the basemat without
the requirement for removal of material
and techniques that can detect and assess
corrosion are of particular interest.
Acoustic (e.g., ultrasonic pulse velocity,
spectral analysis of surface waves,
impact echo, and acoustic tomography),
radar, and radiography appear to have
potential for this application, however,
additional development is required.
The most commonly used type of
foundation for both concrete and steel
nuclear power plant containments is a
mat foundation, which is a flat, thick slab supporting the containment, its interior
structures, and any shield building
surrounding the containment. As
such, the concrete foundation elements
of nuclear power plants are typically either
partially or totally inaccessible for
inspection. Under conditions such as
this the foundation structures are only
accessible for inspection after removal
of adjacent soil, coatings, waterproof
materials, or portions of neighboring
components or structures. As a result,
indirect methods related to environmental
qualification are often utilized
to indicate the potential for degradation
of the nuclear power plant concrete
foundations.21 This is generally done
through an evaluation of the surrounding
medium (e.g., air, soil, humidity,
groundwater, or cooling water).
Remedial Methods
Deterioration of reinforced concrete
generally will result in cracking, spalling,
or delamination of the cover concrete.
Whenever damage is detected,
corrective actions are taken to identify
and eliminate the source of the problem
thereby halting the degradation process.
A remedial measures strategy is
formulated based on the consequence
of damage (e.g., affect of degradation
on structural safety), time requirements
for implementation (e.g., shutdown requirements,
immediate or future safety
concern), economic aspects (e.g., partial
or complete repair), and residual
service life requirements (e.g., influences
action taken). Basic guidance
on the repair of degraded structures is
available.22 Improved guidance, however,
is desired relative to assessment
of defects (e.g., cracks) as well as information
on the performance and effectiveness
of subsequent repairs (e.g.,
durability of repair materials).
APPLICATIONS OF STRUCTURAL RELIABILITY THEORY
If properly designed and constructed,
the concrete structures in nuclear
power plants generally have substantial
safety margins; however, additional investigations
to establish available margins
of degraded structures are desired.
In addition, improved understanding of
how age-related degradation may affect
dynamic properties (e.g., stiffness,
frequency, and dampening), structural
response, structural resistance/capacity,
failure mode, and location of failure
initiation is desired to provide additional
assurance that the current licensing
basis is maintained under all loading
conditions.
Decisions as to whether to invest
in maintenance and rehabilitation of
structures, systems, and components
as a condition for continued service
and risk mitigation, and the appropriate
level of investment, should consider
the nature and level of uncertainties in
their current condition and in future
demands.23 Recent advances in structural
reliability analysis, uncertainty
quantification, and probabilistic risk
assessment make it possible to perform
such evaluations and to devise uniform
risk-based criteria by which existing
facilities can be evaluated to achieve
a desired performance level when subjected
to uncertain demands.24 Reliability-based approaches have been
applied to the nuclear power plant concrete
structures,25 and in evaluation of
the prestress level in concrete containments
with unbonded tendons.26
Degradation effects can be quantified
with fragility curves developed for
both undegraded and degraded components.
Fragility analysis is a technique
for assessing, in probabilistic terms in
the presence of uncertainties, the capability
of an engineered system to withstand
a specified event. Fragility modeling
requires a focus on the behavior of
the system as a whole and, specifically,
on things that can go wrong with the
system. The fragility modeling process
leads to a median-centered (or likely)
estimate of system performance, coupled
with an estimate of the variability
or uncertainty in performance. The
fragility concept has found widespread
usage in the nuclear industry, where it
has been used in seismic probabilistic
safety and/or margin assessments
of safety-related plant systems. The
fragility modeling procedures applied
to degraded concrete members can be
used to assess the effects of degradation
on plant risk and can lead to the development
of probability-based degradation
acceptance limits. This approach
has been applied to a limited extent to
degraded flexural members and shear
walls.27 Figure 4 presents an example
of use of structural reliability theory
to investigate the affect of different assumed
degradation states (i.e., combinations
of loss of steel reinforcement
cross section and concrete cover) on
the probability of failure of a propped
cantilever beam subjected to uniform
loading. Additional work is desired in
this area for the purpose of refining and
applying the time-dependent reliability
methodology for optimizing in-service
inspection/maintenance strategies and
for developing and evaluating improved
quantitative models for predicting future
performance (or failure probability)
of a degraded concrete structure,
either at present or some future point in
time.
CONCLUSIONS
In general, the performance of nuclear
power plant concrete structures
has been very good with the majority
of identified problems initiating during
construction and corrected at that time;
however, aging of concrete structures
occurs with the passage of time that
can potentially result in degradation if
its effects are not controlled. Periodic
inspection, maintenance, and repair are
key elements in managing the aging
of concrete structures. Several areas
have been identified where additional
research would be of benefit to aging
management of nuclear power plant
concrete structures: compilation of material
property data for long-term performance
and trending, evaluation of
environmental effects, and assessment
and validation of nondestructive evaluation
methods; evaluation of long-term
effects of elevated temperature and radiation;
improved damage models and
acceptance criteria for use in assessments
of the current as well as estimating
the future condition of the structures;
non-intrusive methods for inspection
of thick-walled, heavily reinforced
concrete structures and basemats; data
on application and performance (e.g.,
durability) of repair materials and techniques;
utilization of structural reliability
theory incorporating uncertainties
to address time-dependent changes to
structures to assure minimum accepted
performance requirements are exceeded
and to estimate ongoing component
degradation to estimate end-of-life; and
application of probabilistic modeling of component performance to provide
risk-based criteria to evaluate how aging
affects structural capacity.
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D.J. Naus, distinguished research staff member,
is with the Materials Science and Technology
Division, Oak Ridge National Laboratory, Oak
Ridge, Tennessee 37831-6069; nausdj@ornl.gov.
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