Ageing of Nuclear Power Plant’s Equipment. Assessment of Reactor Pressure Vessel Ageing Effect

Abstract

Ageing of materials stands for the change in their mechanical properties, due to thermodynamic imbalance of the initial condition, and gradually bringing the structure to equilibrium in the presence of sufficient diffusive mobility of the atoms. The current paper presents the main mechanisms of degradation of the mechanical properties of the nuclear power plant equipment metal, as well the models for ageing. The ageing effects and ageing indicators typical for the equipment are defined. The ageing process may cause loss of the appropriate functions. It is important to recognize the ageing effects, as well the ageing indicators. Testing methods are defined for every ageing indicator. The paper considers different national approaches to ageing management of structures, systems, and components (SSCs) at nuclear power plants (NPPs).

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Global Journal of Science Frontier Research: H
Environment & Earth Science
Volume 23 Issue 3 Version 1.0 Year 2023
Type: Double Blind Peer Reviewed International Research Journal
Publisher: Global Journals
Online ISSN: 2249-4626 & Print ISSN: 0975-5896
Ageing of Nuclear Power Plant’s Equipment. Assessment of Reactor
Pressure Vessel Ageing Effect
By Galya Dimova
Technical University of Sofia
Abstract- Ageing of materials stands for the change in their mechanical properties, due to
thermodynamic imbalance of the initial condition, and gradually bringing the structure to
equilibrium in the presence of sufficient diffusive mobility of the atoms. The current paper
presents the main mechanisms of degradation of the mechanical properties of the nuclear power
plant equipment metal, as well the models for ageing. The ageing effects and ageing indicators
typical for the equipment are defined. The ageing process may cause loss of the appropriate
functions. It is important to recognize the ageing effects, as well the ageing indicators. Testing
methods are defined for every ageing indicator.
The paper considers different national approaches to ageing management of structures,
systems, and components (SSCs) at nuclear power plants (NPPs).
Keywords: nuclear power plant, ageing, assessment of ageing, long term operation.
GJSFR-H Classification: LCC: TK9152
AgeingofNuclearPowerPlantsEquipmentAssessmentofReactorPressureVesselAgeingEffect
Strictly as per the compliance and regulations of:

Ageing of Nuclear Power Plant’s Equipment.
Assessment of Reactor Pressure Vessel Ageing
Effect
Galya Dimova
Abstarct-
Ageing of materials stands for the change in their
mechanical properties, due to
thermodynamic imbalance of
the initial condition, and gradually bringing the structure to
equilibrium in the presence of sufficient diffusive mobility of the
atoms. The current paper presents the main mechanisms of
degradation of the mechanical properties of the nuclear power
plant equipment metal, as well the models for ageing. The
ageing effects and ageing indicators typical for
the equipment
are defined. The ageing process may cause loss of the
appropriate functions. It is important to recognize the ageing
effects, as well the ageing indicators. Testing methods are
defined for every ageing indicator.
The paper considers different national approaches to
ageing management of structures, systems, and components
(SSCs) at nuclear power plants (NPPs).
An approach has been suggested to study the
effects of ageing of reactor pressure vessels (RPV).
Assessments have been performed on the ageing effects of
reactors as a result of load factors such as radiation and
thermal impact of the neutron fluence, corrosion impact of the
primary circuit fluid, hydraulic and thermal impact of the fluid.
Ke
ywords:
nuclear power plant, ageing, assessment of
ageing, long term operation.
I.
Introduction
toms build up matter and are a source of a great
deal of energy. Atomic energy today is used for
electricity generation, medical and scientific
research, or for exploring of submarine and cosmic
worlds. There are over 441 nuclear reactors in operation
worldwide. A plant’s operating life for a specified
service-time period is justified by the required strength
margin [1]. Normally, the operating design life of nuclear
reactors is 30-40 years [2]. As at April 2022, of all the
reactors in operation, 133 had been operated for over
40 years, while the service life of another 164 had
exceeded 30 years.
Often the owners of Nuclear Power Plants
(NPPs) make decisions to extend the plant life of the
power units: these capacities are the source of various
benefits for
society such as cheap electricity, energy
independence, jobs, knowledge and technological
development. However, in the operation of Nuclear
Power Plants and particularly the older ones, the level of
safety should not be decreased. In Japan, the following
analogy is very popular: nuclear safety culture is
represented as a person standing on the steps of
downward moving escalator. The escalator embodies all
load factors of the equipment, the resulting ageing of
materials and design obsolescence, human errors, i.e.,
those contributors to the reducing of nuclear safety. In
order to maintain one’s position on the escalator, the
person has to make constant efforts, while climbing
upward requires even greater efforts. Continuous
activities are needed to enhance safety culture. In the
energy sector, the problem of ensuring the reliability of
power equipment performance with each passing year
is becoming more and more relevant, as the ageing of
equipment significantly outstrips the pace of
reconstruction and modernization of the operating
capacities. This problem is further complicated by the
absence of a scientifically grounded concept of
technical diagnostics and lifetime determination, as well
as by the inadequacy of traditional non-destructive
testing methods.
The opportunities for design life extension of
nuclear power plants are demonstrated through
analysis, tests and adequate lifetime management for
the expected long term operation [3]. Over the past
decade, a growing number of countries have been
putting the highest priority on the task of lifetime
extension of nuclear power units.
This paper reviews the ageing mechanisms of
the WWER type of equipment, ageing effects and the
corresponding ageing indicators typical for the
manifestation of these effects. Identification has been
made of the control methods and indicators of function
loss of the respective equipment subjected to ageing.
An approach has been proposed for studying
the ageing effects on some of the most important
nuclear power unit components, i.e., the reactor
pressure vessels. Assessments have been performed
on the ageing effects of reactor pressure vessels due to
load factors such as radiation and thermal impact of the
neutron flux, corrosion impact of the primary circuit fluid
and hydraulic and thermal impact of the fluid.
A
Author:
Department of Energy and Mechanical Engineering,
Technical University of Sofia, Sofia, Bulgaria.
e-mails: gtdimova@abv.bg, dimova@tu-sofia.bg
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II. Overview of National Approaches to
Ageing Management of Structures,
Systems, and Components (sscs) at
Nuclear Power Plants (npps)
The technological development level of
individual countries has produced different national
approaches to resolve the problems of nuclear energy
SSCs ageing management.
a) Ageing Management Approach of French NPPs
France has been reported to be one of the first
countries that started dealing with the problems of
ageing equipment at French nuclear power plants.
During the 1980s there were three main lines of activity
and survey [4]: 1) study of the physical process of
degradation with a focus on the radiation embrittlement,
2) the behavior of elements and systems throughout the
ageing processes due to thermomechanical and
hydraulic impacts; corrosion in the primary and
secondary circuits of the power plant, 3) preparation of
reliability analyses and development of methods;
adopting the understanding that operational experience
(OE) serves as a source of data.
At present, France has 56 nuclear reactors in
operation, and 14 ones have gone through a final
shutting-down for decommissioning [4]. France's main
program on the issues of ageing is under the jurisdiction
of the French company Électricité de France (EDF) and
is implemented in three main steps: 1) SSCs with an
impact on the NPP safety, and affected by ageing, get
identified, 2) analyses are conducted on the SSCs
degradation, taking into account the possibility of
maintenance of the facilities, the difficulties regarding
the replacement of obsolete equipment and the risk of
lacking a waste management technology, 3) detailed
reports are written about some components susceptible
to ageing (i.e. the RPV, reactor internals, buildings,
computer and electrical equipment).
The reactor equipment is the hardest and most
complex for replacement. The main degradation
mechanism is neutron ageing of the material, yet ageing
by this mechanism is “well managed” (according to the
French power engineers) except under conditions of
thermal shock. This event could cause brittle fracture of
the reactor pressure vessel material. But such a
scenario is part of the design of the French nuclear
power plants, using special steels with chemical
elements, such as copper and phosphorus that are
naturally resistant to neutron embrittlement. These
elements enable lower temperature values of elastic-to-
plastic transition in the metal. Depending on how low
these reference temperature values are maintained, it is
possible to determine how ageing will be affected and to
predict the performance value at the end of the design
life of the reactor (i.e., after 40 years).
b) Ageing Management Approach of the Hungarian
NPP
In Hungary, there are four WWER type of
reactors (Paks Nuclear Power Station). The plant lifetime
characteristics have been evaluated. The plant life has
been extended by 20 years. A characteristic feature of
the Hungarian approach is that a dedicated Hungarian
regulatory basis has been developed, i.e., the
Hungarian Guideline 4.14 [5], to deal with ageing
issues. The ageing management programs include: 1)
identification of degradation mechanisms and the
affected component (SSC), 2) ageing mitigation and
prevention measures, 3) specifying monitoring
parameters, 4) detection of ageing effects, 5) monitoring
and trending, 6) acceptance criteria of the evaluation
results, 7) corrective actions, 8) feedback, effectiveness
and improvements.
c) Ageing Management Approach of the Spanish NPPs
Spain has seven nuclear reactors. The Spanish
approach regards the SSCs ageing management as a
process that requires periodic re-evaluation and
upgrade. A major source for streamlining the process is
the feedback from operating experience. Many of the
changes to the Spanish NPP ageing management
programs concern the SSCs maintenance activities,
e.g.:
• Preparation and verification of a new guide book on
SSC maintenance activities, with a focus on the
conditions of accessibility to the equipment.
• Training practice in duty tours and walkdowns.
• Improving the identification of structural
components.
The mechanism of flow accelerated corrosion
(FAC) is turning into a challenge to the ageing
management process, as a large part of the carbon
steel equipment necessitates replacement. Another
growing issue is that of inspection and control of some
concealed underground (buried) pipelines, due to
difficulties in using standard tools. Studies are focused
on search of new technological solutions.
d) Ageing Management Approach of the NPPs in the
Czech Republic
The Czech Republic operates six nuclear power
reactors. The Czech methods and criteria for identifying
SSCs within the scope of ageing management require:
• Summarizing the data on equipment ageing.
• Conducting of assessments and documenting
potential mechanisms for properties degradation
that could affect the safety functions.
• Continuous activities to expand the current
understanding of all the dominant mechanisms of
ageing.
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•Availability and adequacy of the data needed for
ageing assessment, including design basis data,
and maintenance and repair data.
•Implementing effectiveness evaluations of the
maintenance and repair programs in terms of
ageing.
•Identifying criteria and indicators for safe operation
over the long-term operation (i.e., operation beyond
the design life term).
•Conducting assessments of the physical condition
of SSCs, including the current safety indicators and
conditions that may limit the operation lifetime.
The ageing management of SSCs important to
safety requires that degradation be controlled in
accordance with specified criteria. Effective control of
the ageing degradation is needed through systematic
assessments of maintenance and repairs, so that this
may result in minimization of ageing trends, and
preservation of the integrity and functional capabilities of
the SSCs.
e) Ageing Management Approach of the NPPs in
Canada
In Canada there are 19 reactors in operation of
the deuterium-uranium unit type (CANDU). In the
equipment screening process, two categories of
components have been identified: critical components
and less critical ones [6]. The following actions are
taken in the course of the ageing management process:
1) assessment of the plant lifetime characteristics of
critical non replaceable equipment (mainly passive
equipment), 2) systematic assessment of the
maintenance actions for critical SSCs through analyses
of the operating states (modes) in which failures occur,
3) condition assessments for less critical equipment
components, and for the remaining SSCs.
The types of critical non replaceable equipment
subject to ageing management include: fuel channels,
steam generators, reactor units, reactor building and
civil buildings, pipelines, turbine generator, pumps and
heat exchangers, electric motors, breakers and cabling
systems, pumps and buildings.
The less critical components and equipment
have been allocated in groups per some typical
characteristic, i.e., commodity groups (pumps, tools).
Each group undergoes specific operability analyses.
The ageing management methodology includes:1)
reviewing the entire operational history of the
component, its design and fabrication in terms of ageing
characteristic features, 2) diagnosing the ageing stress
factors and the mechanism of properties degradation
under all operating modes. Evaluating the component
maintenance in terms of ageing management
effectiveness.
f) Ageing Management Approach of the NPPs in the
USA
The USA have 96 reactors. The components get
assigned to categories based on their significance for
the reliable and cost-efficient nuclear power plant
operation. To this effect the following criteria are used: 1)
ageing effect (potential one), 2) affecting the
component’s intrinsic functions, 3) identifying the
corresponding ageing management activities to ensure
that the expected functions of the components are
supported.
The SSCs assessment is made on the grounds
of NEI 95-10 guideline [7] and is in fact an integrated
assessment of the nuclear power plant and a review of
the time-limited ageing analyses for SSCs covered by
the license. This integrated assessment of NPP consists
of identification of the components’ materials and their
interactions with the environment, the applicable ageing
effects that may impact loss of the anticipated functions,
as well as a lifetime management program needed to
support these functions.
The time-limited ageing analyses contain
qualification for the environmental impacts, fatigue
toughness and neutron embrittlement resistance
analyses.
An element of key importance for the
continuous improvement of ageing management at US
nuclear power plants is the use of OE feedback together
with incorporation of the lessons learned in the ageing
management programs.
g) Ageing Management Approach of the NPPs in the
India
The Nuclear Power Corporation of India Ltd
(NPCIL) conducts the lifetime extension process of the
Indian NPPs in accordance with its own NPCIL
instruction HQI-7005, based on the IAEA Safety Specific
Guide SSG-30 [8]. The main points of this instruction
include, as follows: 1) SSCs screening and ranking as
per level of importance of the NPP safety, 2) ageing
methodology that comprises degradation effect and is
based on the degradation mechanism, 3) evaluation of
the SSC maintenance, 4) inspection of SSCs and
degradation prevention techniques (maintenance,
rehabilitation or replacement).
In terms of material properties degradation, the
equipment has been allocated in four categories:
•Category 1: Major SSCs, of critical importance and
limited lifetime.
•Category 2: Critical SSCs.
•Category 3: Important SSCs.
•Category 4: Other SSCs.
The regulatory system of India does not specify
a limit to the time period for operation of an NPP. The
power plants can continue operation as long as they

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meet the regulatory requirements and satisfy the safety
measures.
h) Ageing Management Approach of the NPPs in China
The first Chinese NPP, Qinshan-1, entered in
commercial operation in 1991. Since then, 15 WWER
nuclear power units have been constructed and
commissioned as well as PWR иCANDU ones.
Safety factors have been checked during each
periodic safety review: 1) documented procedure and
criteria for identification of SSCs impacted by ageing, 2)
list of SSCs incorporated in ageing management
programs, and records that this provides data in
support of the ageing methodology, 3) evaluation and
documenting of each potential degradation mechanism
that may affect the safety functions of SSCs, 4)
broadening of the understanding of the dominant
ageing mechanisms, 5) applicability of data for
degradation assessment including design data,
historical operation and maintenance data, 6)
effectiveness of the operation and maintenance
programs in terms of ageing management of
replaceable components, 7) availability of programs for
timely detection and prevention of ageing, 8)
acceptance criteria and required safety limits for SSCs,
9) informing about the physical condition of SSCs,
including current safety limits and future events (any
events) that may put limitations to the operating lifetime
of the facility.
i) Ageing Management Approach of the NPPs in the
Republic of Korea
The selected criteria for components affected
by ageing and subject to lifetime extension involve a
number of standards and normative documents
including regulations for periodic safety reviews (PSR),
US regulatory documents [7], and definitions of quality
class implemented in Korea:
•Safety related components (Quality Class „Q“).
•Non-safety related components the failure of which
may affect safety functions (Quality Class „A“).
•Other components.
The current physical condition and level of
degradation of the SSCs are evaluated by means of
referring to the design and manufacturing data, taking
into consideration the data from testing, operation and
maintenance, in accordance with the applicable
standards.
The ageing management review produces
analyses of whether the selected properties degradation
mechanisms (ageing mechanisms) can affect SSC.
These mechanisms are evaluated as follows: 1) It is
determined if the ageing mechanism found for a given
item forms part of the 17 mechanisms as specified in
ASME Boiler and Pressure Vessel Code [9], 2) each
ageing mechanism is identified, 3) the frequency and
conditions of the occurrence of each ageing mechanism
get identified, 4) determination is made of the type of
mechanism(s) applicable to the examined component,
5) consideration is given to the availability and
applicability of any operational experience.
j) IAEA Documents of the Ageing of NPP Equipment
The document IGALL Ageing Management of
Nuclear Power Plants [10] provides guidelines on the
potential content of an integrated ageing management
program, as well as on the assessment of the ageing
management program effectiveness. The NPP self-
assessment of the ageing management measures shall
include, as follows: 1) implementation of all in-plant
measures for safety assessment, such as PSA periodic
conduct, 2) conduct of external oversight -on behalf of
the regulatory authority, as well as on international level
(SALTO review missions of the IAEA), 3) comparing, on
a periodic basis, the ageing management activities
against those implemented on other NPPs
(benchmarking process). It is assumed that the ageing
mechanisms of the same type of equipment are the
same on the different NPPs. However, an ageing effect
may be manifested on one NPP, and not manifested on
another one, or occur at a later stage of the operation of
the particular equipment. In view of this, benchmarking
against the practices of other NPPs assists in the
prevention of ageing.
IGALL Ageing Management of Nuclear Power
Plants [10] contains tables covering all types of
materials on an NPP, all types of components, intrinsic
degradation mechanisms and ageing effects.
The unified procedure Nulife, or Verlife [11] is a
technical document (ТеchDoc) of the IAEA and it
provides a methodology for: 1) assessment of the
residual lifetime and integrity of components and piping
of NPPs with WWER type of reactors in the course of
their operation and in terms of defects caused by ductile
and non-ductile fracture, fatigue and mechanical
corrosion damages as a result of their operation, 2)
assessment of the indications found during in-service
inspection (ISI) of components and pipes, 3) preparation
for reports from the periodic safety review during an
NPP operation, in the part regarding the residual lifetime
of equipment, 4) management of modifications of NPP
equipment residual life.
III. Ageing Effects Study Methodology
Technical disciplines have been emerging
based on requirements for failure and defect prevention
and ageing management of mechanical and electrical
system for plant life extension [4].
Failures and defects of equipment and pipelines
occur when a limit condition has been reached. Limit
conditions are attained in the following circumstances:
1) upon reaching of unacceptable residual changes of
form due to plastic deformations, corrosion, mechanical
or erosion wear, 2) upon the emergence and growth of

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discontinuities, 3) when the service life characteristics
have reached their ultimate limit values, for example the
acceptable number of load cycles.
The approach to ageing effects study shall satisfy the
following main requirements:
•Protect the equipment from ageing;
•Monitor the consequences of ageing;
•Compensate the consequences of ageing;
•Improve the equipment control programs in the light
of new knowledge accumulated including the
programs for surveillance specimens’ analysis.
•Effectiveness assessment of the testing methods
and the programs.
a) Preparations: Data Acquisition
A thorough and extensive data acquisition is
effected through review and analysis of the design and
manufacturing documentation of each component,
resulting in systematization of: 1) datasheet and design
data about the facilities, 2) design changes and
modernizations undertaken, 3) provisions of the
normative documents, 4) operational history and testing
data, 5) maintenance documentation, 6) strength
analyses data, 7) data about compatibility with other
components and systems, 8) results from in situ
inspections (control), 9) data on the implemented
operating modes, hours, number of load cycles,
hydraulic tests, etc.
b) Strength Analyses
c) Selection of Sample Components
Ageing affects all the SSCs of a nuclear power
plant. Naturally, it is not possible to subject to systematic
survey for ageing effects all their thousands of
components. Screening of the components due for a
more extensive analysis of ageing is affected on the
grounds of various factors, e.g.:
•The time in which they have been in operation.
•Number of the strength cycles.
•Conditions of operation, especially those conditions
that are most relevant to the crack formation.
•Which components can be renovated through
welding.
Components are allocated in groups on the
basis of similarity signs. These signs may differ, but
most frequently components are grouped by functional
identity (e.g., the steam generators group). Other type of
grouping may be founded on the results from testing,
e.g., a group of components with defects that exceed
certain size.
From the group, a component is selected
because it is in the most unfavorable position in terms of
ageing: for example, it has the worst physical condition,
or it is the hardest to accessible for control purposes, or
it has been subjected for the longest time to the
aggressive influence of environment. From the entire
group of components, one sample component can be
selected to be used for an in-depth analysis of the
ageing trend. After conducting these analyses, the
results can be applied to the other components in the
group. After a sample component has been selected
and its condition analyzed, it is assessed whether there
is a need to expand the scope of inspections and
monitoring, carried out up to that point, so as to cover
the entire group of components. In the event of
obtaining satisfactory results for a typical component, it
is not necessary to extend the scope of the inspection.
While if the results of the control (testing) are not
satisfactory according to one or more criteria applied,
then an increase in the scope of the ageing control may
be recommended.
For each component, data is collected. The
data collection for each component is systematically
supplemented with results from regular non-destructive
and other testing, component maintenance data, and
information on high stress potential areas.
d) Regular and Extraordinary (Additional) Studies of
Characteristics of the Mechanical Equipment
The regular studies due for each component
have been described in the technical specifications for
the operation of the nuclear power unit.
Additional measurements include, as follows: 1)
precise measurement of the mechanical properties by
means of a kinetic penetration method (kinetic hardness
The performance of strength analyses is a
mandatory part of the ageing management process
[12]. All the relevant information on structural materials,
geometry and design characteristics, the rest of the data
from the design documentation and the components’
datasheets serve as input data for the strength
analyses. The cyclic fatigue effects, caused by the
operational fluid need to be considered in the
calculations. The current operational modes, all data
about defects and non-conformances regarding the
design parameters also need to be taken into account in
the strength analyses. The average loads spectrum over
the past 10 years of the NPP operation shall be used as
a model for the future annual load. The calculations may
also consider the rest of the loads. Based on the results
of the strength analyses, it has been found that in the
most stressed areas the metal can potentially be
exposed to the highest level of ageing resulting in
degradation of physical-mechanical properties, due to
ageing factors such as thermal deformation and low-
cycle fatigue. Occurrence of a crack may be expected in
the zones of highest stress (load). Therefore, in those
areas where the strength analyses indicate a potential
for failure, special emphasis is placed on testing and
non-destructive testing, for example, in the places of
welded joints.

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method), 2) ultrasonic testing of the welds integrity by
means of the phased array technique, 3) on-line
measurement of the decreasing wall thickness of metal
facilities using combined electromagnetic-acoustic
methods for selected carbon steels of pipelines subject
to erosion-corrosion. This method enables 3-D scanning
of the object, in order to detect cracks and wear, and
permits the identifying and localization of the area of
maximum metal wear.
e) Evaluation of the Adequacy of Ageing Management
Activities
Ageing management activities include testing,
monitoring, control, feedback and operational
experience implementation, etc. These measures are
apparently sufficient if the component is in good
condition. On some NPPs, it has been established that
for some components, for which control/monitoring is
not required as per the unit's technical specifications, no
tests have been conducted at all during the years of
operation. This fact points to serious gaps in the
maintenance and repair system, incompatible with the
management of ageing processes:
1) equipment has been found not covered by
the maintenance and repair measures (polar crane and
refueling machine, RPV supports), 2) equipment that is
on the borderline between two systems and has not
been included in the scope of the maintenance and
repair programs, 3) no regular periodic measurements
have been made of the mechanical properties (strength,
hardness) in areas with potential for failure.
In such cases, additional examinations of the
mechanical properties of the metal are prescribed and
carried out, in addition to the regular surveys, for
example, a hardness testing is prescribed. The
adequacy of the maintenance and repair measures is
evaluated in terms of the activities described above.
f) Identifying the Degradation Mechanisms of the
Mechanical Properties
The mechanisms of mechanical properties
degradation are identified for each commodity group
[10,13]. The ageing effects attributable to each
mechanism are determined. Determination is also made
of the ageing indicators through which the effects are
manifested. The testing and diagnostic methods are
defined by means of which the ageing indicators are
monitored. These methods get described in ageing
management programs for each commodity group.
g) Evaluation of the Effectiveness of Ageing
Management Measures
The evaluation of the effectiveness of ageing
management measures shall take place on a regular
basis. The information items listed below serve as input
data for evaluations [14, 15]
•The component condition, including the defects
manifested in the facilities.
•The degradation mechanisms, i.e., those already
defined and the potential ones.
•Whether the ageing effects are typical
(characteristic) of these mechanisms.
•Whether the ageing indicators have been correctly
identified.
•Whether the testing methods (control and
monitoring) are sufficiently sensitive to capture the
changes in the ageing indicators.
•Whether the periodicity of testing is adequate. If
discontinuities have been found, the interval
between inspections has to be shortened.
•Whether the scope of the inspected equipment is
adequate.
•Let’s take a closer view of the process of ageing of
materials.
IV. Ageing of Materials and
Degradation of Mechanical
Properties
Ageing of materials stands for the change in
their mechanical, physical and chemical properties, due
to thermodynamic imbalance of the initial condition, and
gradually bringing the structure to equilibrium in the
presence of sufficient diffusive mobility of the atoms.
Investigations of the ageing processes during NPP
operation comprises activities such as
•Development of methodologies and instruments to
diagnose the parameters of NPP in-service
equipment.
•Assessment of the ageing effects on the operability
of equipment in view of making corrections to the
scope and periodicity of outages, maintenance,
tests and inspections.
•Development of methodologies for express analysis
of failures, damages and defects of components of
equipment, and introducing those methodologies in
the operational practice.
•Establishing an NPP information system (a
database of knowledge and expert system).
a) Corrosion
Corrosion is the process of metal failure as a
result of chemical or electrochemical interactions of
metal with the surrounding environment [16]. The cause
of corrosion is the thermo-dynamic instability of the
system, composed of the metal and components of the
environment. The capability of metals and alloys to resist
corrosion impact of the environment is contingent on the
rate of corrosion under the given conditions. The
following serve as quantitative indicators of the rate of
corrosion

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•The time until the occurrence of corrosion
outbreaks.
•The number of corrosion outbreaks for a given time
period.
•The metal thickness decreases per time unit.
•The change of metal mass per surface unit and time
unit.
•The change of any indicator of mechanical
properties such as strength, plasticity, electrical
resistance, etc.
b) Erosion
Erosion of the walls of equipment is caused by
particles of various origin such as particles of metal
corrosion products, sand, silicates, water drops, etc.
The erosion process evolves through brittle or plastic
fracture depending on the temperature.
1. Under normal temperature conditions in the plastic
metals, erosion dissociation of metal occurs as
results of plastic deformation on the surface. With
brittle materials, erosion takes place through surface
degradation in the form of cracking.
2. High temperature erosion is associated with the
release of composite material –metal alloy and
brittle surface oxide. The oxide layer on the metal
surface may modify the process mechanism
depending on the layer thickness. If the oxide layer
is thin, the prevalent mechanism is associated with
metal creep (elastic-plastic area). Upon the oxide
layer reaching a critical thickness, the dominant
mechanism is that of brittle erosion fracture.
The temperature and the characteristics of the
force impact of the particles are the erosion determining
parameters. The speed of oxide formation is dependent
on temperature and, therefore, the same applies to the
oxide layer thickness within a given timeframe. The force
effect of particles is characterized by the time intervals
of particles impacting on a specified point on the metal
surface.
c) Ageing Effects Due to Corrosion-Erosion Processes
The corrosion-erosion processes that are typical
for NPPs, type WWER consist in corrosion degradation
of materials, followed by erosion wear under the impact
of the fluid flow rate.Factors affecting the process
include: 1) fluid composition, 2) velocity and
temperature, 3) the component material and geometry,
4) the active stresses, 5) the periodicity of surface
moisturizing / drying. Localized corrosion affects steam
generators and reactor sealing surfaces, pressurizers
and emergency core cooling system (ECCS).
Intergranular corrosion can be observed on reactor,
steam generators, corrosion fatigue –on steam
generators and pressurizers. Stress corrosion affects
reactors, steam generators, pipelines of the pressurizer
systems and ECCS piping.
Stresses will lead to a considerable change in the metal
electrode potential. Tensile stresses (tensions) shift the
electrode potential to the negative side, while the
compressive stresses shift it to the positive side. The
stretched sections act as anodes with regard to the rest
of the metal and degrade (dissolve) most intensively.
The corrosion-erosion processes decrease
pipeline wall thickness. There is increased probability of
pipeline rupture and leaks of coolant. The change in the
geometric dimensions of the pipeline walls leads to a
change of the internal stresses. The system’s failure rate
increases on account of material degradation. The
corrosion-erosion processes are the cause for loss of
tightness of the pipeline systems. Thus, the normal
operating conditions are compromised. Abnormal
operation of the heat exchangers occurs as a result of
the rupture of heat exchanging tubes. The general
radioactivity levels increase due to activation of the
corrosion products.
One of the ageing indicators is the pipeline wall
thickness and it is measured periodically. The corrosion
rate is inspected on a periodic basis. Tests are
performed to identify presence of number, type, location
and growth of surface defects, pits, and blow-holes in
metal, and percentage of wear of the wall thickness of
heat exchanging tubes. Conduct regular ultrasonic
thickness measurement testing of walls and bends. The
area surround-ding the weld joints is monitored.
Monitoring is performed of the water chemistry
of the fluid inside the pipeline. The water chemistry
regime is analyzed and maintained. The radioactivity
indicators are measured. Corrosion evaluation for
presence of sludges or deposits is to be performed of
the critical areas. In-service inspection of metal is
performed (visual, penetrant, eddy-current and
mechanical testing). To prevent intergranular corrosion,
visual inspection, surveillance specimens testing,
penetrant, ultrasonic and hydraulic testing are
undertaken.
d) Neutron Embrittlement
The operating conditions of the reactor pressure
vessels metal are characterized by intensive neutron flux
under high temperature and pressure conditions [17].
Being particles of small mass and great energy,
neutrons easily penetrate the crystal lattice of the reactor
pressure vessel. There are two major mechanisms of the
interaction between neutrons and the particles of
materials:
1. The collision between neutrons and the lattice
atoms causes dislocations within the crystal lattice;
neutrons may either transfer their energy to atoms
through elastic impacts, or serve as the source of
charged particles formation. Such processes will
impair the correct position of atoms within the metal
crystal grid and this will lead to defects formation. In
Regarding the bends in the pipeline system:

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initially displaced from its balanced position may
be followed by a cascade of displaced atoms.
2. Radiation impact largely facilitates diffusion of the
ingredient’s atoms, which is another important
cause for alloy embrittlement. Moreover, as results
of vacancies merging in those diffusion processes,
additional pores may form in the metal, which can
result in noticeable changes in the shape of the
structure.
The density of radiation defects depends on the
type of radiation, its parameters and the nuclear-
physical characteristics of the material. The spot defects
that occur –vacancies, internodal atoms, embedded
atoms at sufficiently high temperature can recombine,
migrate to body or surface directed leakages
(dislocations, grain boundaries) form radiation stacking
faults in the shape of pores and dislocation nodes. The
irradiation of metal with fast neutrons results in
microscopic areas of structural damages, and with high
concentration of spot defects. Due to irradiation, the
creep (yield) stress limit of steel may grow up to twice
fold, while the strength limit increases to a lesser degree
–the two limits come closer and metals harden while
also losing plasticity. Current knowledge of radiation
degradation assumes that the occurring defects may
lead to material hardening either directly via interaction
with the dislocations, or indirectly –through the
changing kinetics of metallurgical reactions leading to
phase drop. These effects harden the material and are
dependent on neutron fluence density. The main effect
of radiation degradation of metals consists in the highly
limited number of active slip planes, and increased
number of dislocations moving across the slip planes.
This highly localized movement affects the process of
local degradation in the peak of the crack. Determining
the transition from elastic to brittle state or the evaluation
of the shift of the brittleness critical temperature ∆due
to the neutron fluence , may be performed through
experimental testing of surveillance specimens, or it can
be assessed numerically through the neutron fluence .
Neutron embrittlement is expressed in radiation
brittleness temperature ∆shifting in the direction of
higher temperature values. Numerical assessment of
neutron embrittlement of the reactor vessel metal is
carried out using norms and standards of the country
manufacturing the reactor equipment.
Due to the neutron diffusion, near the peak of
the crack a circle section of embrittled metal forms, as is
shown on “Figure 1”. The embrittling action of metal
neutrons is dependent on their density of distribution
[17].
Figure 1: Material Embrittlement Around the Crack Peak
e) Effects of the Chemical Composition of Steels on the
Embrittlement
The steels used for NPP’s equipment are of
ferrite-perlite type. The elevated levels of the elements
nickel Ni and manganese Mg in reactor steel grades
enhance embrittlement due to the formation of nickel-
manganese-silicon Ni-Mn-Si clusters (dislocation
nodes). “Figure 2” shows photos of microscope
examination of samples with various weigh percentage
of Ni, subjected to neutron fluence irradiation [18].
case of sufficiently high neutron energy, the atom

0.02 %Ni 0.82 %Ni 1.59 %Ni
Figure 2: Formation of Dislocations in Samples of Varying Weght Percentage of Ni (0.22 Cu – xNi - 1.6 Mn) at
Temperature 290- 320 ⁰C
During irradiation the structure of materials
containing copper Cu changes and Cu-enriched
clusters form. The Cu-nucleus stays in the middle of the
formation, while theelements nickel Ni, manganese Mn,
silicon Si accumulate in the outlying sections.
These formations disrupt the correct structure of
the crystal lattice. Under the impact of the operating
temperature of 320 ⁰C, higher phosphorus content will
result in thermal brittleness of metal following a
mechanism based on the phosphorus segregation at
the interphase boundaries and the grain boundaries.
Radiation embrittlement is determined by the formation
of dislocation nodes. The occurrence of these defects
results in: 1) facilitating the emergence of cracks and
development of micro crack on account of the active
stresses, 2) additional micro stresses begin to act in the
grain bodies, 3) increased probability of formation of
dislocation aggregates at the barriers where micro
cracks form. The neutron embrittlement mechanism is
associated with the segregation of phosphorus at the
inter-phase boundaries of the carbide matrix and at the
grain boundaries, as a result of which their strength
diminishes.
f) Thermal Ageing
In NPP facilities in operation undergo the impact
of high temperature values. These working medium
factors may cause thermal ageing of the base metal and
weld metal in terms of alteration of their mechanical
characteristics [19]. Thermal ageing is associated with
displacement of atoms in the lattice of the crystalline
structure; it is mainly dependent on temperature and the
time period over which the metal has been exposed to
its effects.
Regarding the reactor vessel materials, neutron fluence
causes both thermal ageing and radiation ageing of the
metal. The thermal ageing mechanism has been
determined by the carbide’s formation process. In the
course of thermal treatment, carbon bonds in stable
carbides do not change under the operating
temperatures over the whole service life of the materials.
Upon carbides emergence and as their amount grows,
the material hardens and, as a result of this, it also
becomes brittle.
The thermal ageing affects the reactor pressure
vessel, barrel, core baffle, and reactor guard-tube bank.
Factors affecting the process include: 1) fluence values
and direction, 2) the chemical composition of materials.
The elevated content of nickel Ni and manganese Mn
enhance thermal embrittlement due to the formation of
clusters in the radiation environment, while the content
of silicon Si reduced embrittlement.
The loss of functions due to the thermal ageing
is observed. There is a growing probability of brittle
fracture of materials especially for the welded joints of
the reactor vessel located opposite the reactor core. As
a consequence of radiation swelling of metal there is
growing probability of shape changing of components
(core barrel); this will lead to altered load bearing
capacity of the structure.
Both neutron and thermal ageing can be
presented with the embrittlement function (,) –
the shift of the critical brittleness temperature, which is
depended from the values of neutron fluence F and time
t [9,10]. The value of the shift (,) needs to be
within the acceptable design limits [8]. The parameters
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monitored are presence of number, type, location and
development of surface and internal discontinuities. The
neutron flux values following each fuel cycle are
recorded. In-service inspection of metal is performed
(visual, penetrant, ultrasonic, eddy-current and
mechanical testing). The mechanical characteristics are
studied periodically, including the variation of the shift
(,) of surveillance specimens from the reactor
vessels. Thermal hydraulic analyses are conducted, as
well strength analyses. Low-leak schemes of core
refueling are used. Visual and measurements inspection
of the core barrel are performed. The barrel geometry
dimensions are monitored.
g) Material Fatigue
Engineering structures are subjected to
pulsating (cyclic) loads [20, 21]. Under the in-fluence of
cyclic loading, it is difficult to notice any progressing
changes in the structure of the material. Destruction
happens suddenly, without any noticeable signs of
imminent occurrence. Moreover, in times of “relaxation”
when stress stops acting, defects do not disappear –
they accumulate and are irreversible. Fatigue can be
described as a process of gradual degradation,
composed of sub elements, such as: 1) the process of
crack emergence, 2) crack growing to a size when its
further progress is rapid and unstable. It is assumed that
a crack originates as a result of the movement of
dislocations, which generates thin sliding planes on the
surface of the crystal lattice.
h) Ageing Effects Due to Fatigue
Cyclic fatigue affects all the main equipment
pieces of the primary circuit (reactor, steam generator,
main circulation pipeline, main coolant pump). Low
cycle fatigue affects the secondary circuit equipment
(turbine, demineralizers, separators).
Factors affecting the process include: Number
N of the work cycles, amplitude оf the deviation of the
stress. Loss of function of the equipment due to fatigue
is observed. There is increased probability of fatigue
degradation of materials. Subsequent change in the
load-bearing capability of structures is expected.
The number of load cycles is monitored for the
different operating modes. Monitoring is performed of
the following parameters: presence of number, type,
location and development of surface, below surface and
internal discontinuities. The fatigue accumulation factor
is periodically calculated. The load cycles are registered
and monitored for the different design operating modes.
A register is maintained of the number of loading cycles.
Surveillance specimens are tested. In-service inspection
of metal is performed (visual, penetrant, ultrasonic, eddy
current testing) [22, 23]. Hydraulic testing is
implemented. Fatigue analyses are performed.
i) Wear
Throughout the stage of alignment mutual
changes occur in the macro- and micro-geometry of the
working faces, and products of wear and oxidation are
formed. The working faces wear rather intensively during
this stage. Gradually, wear weakens and stabilizes to a
stage of normal operation wear. Once the energy limit
has been exceeded, the wear value progressively
increases, the components functioning deteriorates and
the need of repair arises. The following factors
determine the level of wear in friction: 1) physical,
chemical and mechanical properties of the surfaces
subjected to friction,2) combination of materials for the
working surfaces, 3) interactions of the working surfaces
with the environment, 4) clean processing of the friction
surfaces, 5) type of friction (dry, boundary, semi liquid,
liquid), 6) values of the normal pressure and the velocity
of working surfaces one against the other.
Of the large number of wear types on the
working surfaces of machine parts, major importance is
attached to abrasive wear in the presence of grease,
because the wear products that invariably arise from the
machine components friction are oxidized and turn into
a sort of abrasive materials and it is rather complicated
to clear lubricants from the component surface. Friction
wear is one of the major contributors to the gradual loss
of operability of mechanical elements. Therefore, the
consideration of factors affecting the level of wear of
machine parts during design and operation of
mechanical systems is one of the main tasks for
ensuring the reliability of the working mechanical
elements in Nuclear Power Plant.
j) Ageing Effects Due to Wear
Wear affects hydraulic snubbers, sealing faces,
fixing elements, internal parts of cylindrical vessels and
pipelines. A photo of wear is presented on “Figure 3”.
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Multiple studies have demonstrated that the
process of gradual loss of functionality of components
in operation can be subdivided in three stages: 1) stage
of alignment, 2) normal operation stage, 3) wear, caused
by facilities’ normal operation.

Figure 3: A Photo of Wear on Rotating Part, it Can Be Seen a Defect in Metal
Erosive wear is an issue for pipeline operation in
the turbine hall. Defects of erosive nature occur at
pipeline bends, and also in pipeline sections
downstream of throttle and control valves. The cause for
such defects is the presence of two-phase medium in
the pipes and development of cavitation processes.
Plastic deformation affects steam generators,
pressurizers, pipelines, of the pressurizers’ system, main
circulation pipelines and ECCS.
Loss of function of the equipment due to wear
can be observed in a case of hydraulic snubbers’
degradation. The snubbers are uncapable to performing
its protective functions in case of strong vibrations, or an
abrupt displacement of equipment caused by seismic
loads. Regarding pins (their cylindrical part) and pin
sockets – their fixing functions is impaired and it is
probable that fixing will not be tight enough.
Erosion-corrosion wear processes decrease
pipeline wall thickness. There is increased probability of
pipeline rupture and leaks of coolant. The load-bearing
capacity of the affected structures is changed.
Visual testing is performed to identify presence
of number, type, location and development of any
surface discontinuities. The pipeline wall thickness is
measured periodically.
V. Ageing Effects of Reactor Pressure
Vessels
This part of the paper reviews ageing effects of
reactor pressure vessels (RPVs).
The underlying factors for selecting these items to be
subjected to ageing assessment are as follows:
1) The operating life is greater than 30 years in a row.
2) Large number of the strength cycles.
3) Operating environment conditions: fluid pressure of
17.8 МРа; fluid temperature 20 ÷ 330 ⁰ С; the fluid
flows at high speed.
The subject of assessment was the reactor
pressure vessel metal, and the RPVs type WWER 1000 –
B 320, thermal power of 3000 MW. Two power units were
the subject of survey: unit “a” and unit “b”. The survey
period covers 30 years.
The RPV metal is subject to the following
degradation mechanisms due to the operating
environment factors such as neutron embrittlement,
thermal ageing, embrittlement due to the presence of
discontinuities, fatigue, and erosion-corrosion wear [13].
a) Study of the RPV metal ageing caused by neutron
embrittlement and thermal ageing mechanisms
Subject of study are the welded joints metals of
the RPVs of two type WWER reactors (provisionally
referred to as “a” and “b”). The reasons justifying the
choice of the specific areas for the study is that areas
with degradation potential are identified in these RPV
points. The welded areas have different metal structure
(area of base metal, thermal impact area, and welded
metal area) and the structural non uniformity is one of
the main causes for the occurrence of discontinuities.
The tensions active in the metal and resulting from the
operating conditions are not the same for the bimetallic
areas, which leads to the growth of discontinuities. The
welded joints located opposite the reactor core are
subject to the embrittlement action of high neutron
fluence. Discontinuities were found at the places of
welding; their evolution has been traced over the years
of reactor operation.
The embrittlement process depends not only on
the chemical composition of the alloys, but also on the
values of neutron fluence, operating temperature and
running hours, which can be expressed as shown below
[11, 19]: =0+(,)
(1)
The value added to the temperature (,)
shift has two components: one of the components is
due to the neutron fluence (), and the other one is
due to the thermal embrittlement ().
(,)=()+()+ (2)
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• 0 is the metal initial critical temperature that
corresponds to non-irradiated condition;
• is the metal critical temperature following a period
of irradiation;
• () is the metal critical temperature shift,
resulting from thermal ageing;
• t running hours of the reactor metal;
• () the critical temperature shift, resulting from
neutron irradiation due to the neutron fluence F;
• Fis the neutron fluence of neutrons whose energy
exceeds 0.5 MeV, hitting the pressure vessel;
• 0=(1022 n)/m2 is a standardised factor;
• is the radiation-induced embrittlement factor;
•  – double standard deviation of ();
•  - is the embrittlement critical temperature shift
at t=;
• ,,– material constants;
• Ni, Mn, Cu and P represent the chemical elements
concentrations in the metal composition, [weight
units];
• =72.1022 , is a standardised factor.
()=(/0) (3)
for base metal: m=0.8 AF=1.45 [⁰C]
for weld metal: m=0.8 AF= 1.exp2.. [⁰C]
.=+3. if +3.0
or .= 0 if +3.< 0
1= 0.703;2= 0.883;3= 3.885;
()=󰇣 +.󰇡
 󰇢󰇤.󰇡
󰇢 (4)
The values of the quantities ∆T inf,bt, tОТ for the
pressure vessel metal are summarized in “Table 1”.
Table 1: Values of the Quantities ∆Tinf , bt, tОТ
Metal Table Column Head
∆Tinf, ⁰C
bt,
⁰C tОТ,hours
Base metal 18 26.2 32 700
Weld metal,
Ni>1,3 % 18 10.1 23 200
Weld metal,
Ni<1,3% 18 26.2 32 700
Two methods are known for determining of the
critical temperature shifts (,).
The first method is a theoretical one -
calculations using certain numerical models adopted in
normative and methodological documents. The second
method is a practical one - through analysis of
surveillance specimens’ material. The input data for
assessments of the ageing effects are
• Datasheets with the composition of the reactor
pressure vessels (Passport data).
• Data of the fluence on the RPV in the course of each
fuel cycle (campaign).
• Data from NPP logbooks about the running hours in
each fuel cycle.
• Data from the surveillance specimens testing.
The theoretical method for analysis of the
critical temperature shift (,) is based on
calculations. The values of fast neutron fluence with
energy greater than 0.5 MeV, reaching the inside of the
RPV wall are monitored through the neutron detector
readings positioned around the reactor pressure vessel.
Data sampling is performed once a year.
The practical (experimental) method for analysis
of the critical temperature shift (,) is based on the
results from surveillance specimens impact strength
tests. Calculations were made of the embrittlement
critical temperature (,) on two RPVs with WWER-
1000 reactors (referred to as “a” and “b”).
Figure 4: Function () for Base Metal BM and Weld Metal WM of the RPVs, units “a”, “b”
0
5
10
15
20
25
30
1234567891011121314151617181920212223
∆Tk(F) , [ºC]
Neutron fluence F. 1022 [n/m2]
BMa,BMb
WMa
WMb
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The critical temperature shift ∆(), caused by
the neutron fluence on base metal and weld metal is
shown on “Figure 4”.

The critical temperature shift values ()
grow proportionately to the increase of the fluence F
values, both in the base, and the weld metal. The
() curves for base metal (BMa and BMb) coincide
because  does not depend on the chemical
composition of the base metal. Difference can be
observed in the neutron embrittlement rate of weld metal
(WMa and Wmb); greater embrittlement rate was found
for WMa. Regarding RPVb, the curves ()for base
metal, BMb, and weld metal, WMb, almost tally.
The critical temperature shift resulting from
thermal ageing () of base metal and welded metal
(BMa, BMb, WMa, WMb), is shown in “Figure 5”.
Figure 5: ()function of the Time t for base metal and weld metal of the reactor pressure vessels of units “a” and
“b”
The curves’ trend (behavior of the curves) for
the thermically induced part () for base metal
(BMa, BMb) anticipates the trend of the curves for weld
metal (WMa, WMb). A peak can be observed for the
thermal embrittlement values over a period of 2-5 years,
following which the thermal embrittlement drops sharply.
After the first 10-11 years of operation, the function
() has almost unchangeable value, and this trend is
preserved over the further operating period of the metal,
both welded and base metal.
The sum of the critical temperature shifts caused by
neutron and thermal ageing is:
(,)=()+() (5)
The function (,) for base metal and weld
metal is shown in “Figure 6”.
Figure 6: Embrittlement (Neutron and Thermal) for base Metal BM and Weld Metal WM The (,) Function of T
time
The thermally induced embrittlement is
dominant in the first 10 years of an NPP operation, after
which the neutron embrittlement is prevalent. In the
beginning of NPP operation the embrittlement rate of
base metal BMa, BMb prevails over that of the welded
metal WMa, WMb. This is followed by process reversal,
i.e., prevalence of weld metal embrittlement.
A comparative analysis was conducted of the
critical temperature shift values (,). Comparison
was made between 1) (,) as calculated with the
fluence value of the neutron detectors, and 2) (,)
obtained on the basis of experimental data from the
surveillance specimens (SS). The results are
demonstrated on “Figure 7”, a) base metal BM and b)
weld metal WM.
0
5
10
15
20
25
30
35
40
45
12345678910 11 12 13 14 15 16 17 18 19 20 21 22 23
∆Tk(t) , [ºC]
Time t, [years]
BMa
WMa,WMb
BMb
0
5
10
15
20
25
30
35
40
45
50
12345678910 11 12 13 14 15161718192021222324 25 26 27
∆Tk(F,t), [ºC]
Time t, [years]
BMa
WMa
WMb
BMb
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Figure 7: Function (,) of Time t for a) Base Metal BM (above) and b) Weld Metal WM. Averaged Line based on
the Experimental Data from the Surveillance Specimens for a) Base Metal BM and b) Weld Metal WM
For base metal – the calculated results are
higher than the experimental ones. The (,) function
trend exceeds the values of the experimentally obtained
data from the surveillance specimens of unit “a” base
metal BMa, contrary to the case with base metal BMb.
For weld metal WMa, the trend curve (,),
based on the calculations is high and quickly grows
proportionally to the operating time (running hours)
increase. The trend curve (,) for WMb (almost)
coincides with the averaged line based on the
experimental data.
The RPV metal resistance to neutron and
thermal impacts is evaluated by means of the values of
(,) - cold embrittlement critical temperature. With
the increase of the neutron fluence value F and of the
accumulated running hours, t, the values of (,)
become higher. The values of (,) shall be regularly
compared against the normative and the design
specifications. This a main requirement of the technical
specifications for safe operation of the reactor
equipment. With (,)< Тmargin (margin value),
resistance of the reactor pressure vessel metal to
neutron and thermal impacts has been achieved.
1) The calculated and the experimentally obtained
values of (,) show that the critical
embrittlement temperature shift is less than Тmargin
=57 ⁰C. Therefore, the requirement for resistance of
the RPV metal to neutron and thermal impacts has
been satisfied for units “а” and “b”.
2) The curve (,) trend for welded metal with
higher content of Ni exhibits the greatest dynamic. It
can be concluded that these welded joints (unit “a”)
are the most critical element of the reactor. This
inference is confirmed by the match between the
calculated and the experimentally obtained data for
(,).
3) Regarding base metal, a good match was obtained
between the calculated data for units “а” and “b”.
The experimental data from the surveillance
specimens testing demonstrated lower values for
the lifetime characteristics (,).
0
5
10
15
20
25
30
35
40
45
50
12345678910111213 14151617 18192021 222324252627
∆Tk(F, t), [ºC]
Time t, [years]
SS_BMa
SS_BMb
BMa
BMb
0
5
10
15
20
25
30
35
40
45
12345678910 11 12 13 14 15 16 17 18 19 20 21 22 23
∆Tk(F, t), [ºC]
Time t, [years]
WMa
WMb
SS_WM
b
SS_WM
a
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b) Study of the RPV ageing metal due to corrosion-
erosion wear
The RPV metal, both on the inside and on the
outside surfaces can be tested (controlled) through
scanning using a remote system for visual inspection.
The periodicity of this activity is once in 4 years, as
specified in the technical specifications (for operation of
the nuclear power unit). If needed, this period can be
shorter. The visual inspection method enables detecting
and diagnosing surface discontinuities on the RPV inner
surface. The controlled parameters are: presence or
absence of discontinuities, their type, size and location
[23]. To examine metal for corrosion-erosion (ageing)
wear, remotely operated visual inspection is
implemented. An underwater camera system is used to
examine the RPV; A special software serves for storing
data on the location of indications, sizing, comparing
with previous data.
The input data for the studies are the
parameters of discontinuities found on the inner surface
of the RPV and include, as follows: 1) type of
discontinuities as classified according to a standard, 2)
discontinuities’ location and coordinates in the metal,3)
size of discontinuities, 4) orientation of the
discontinuities.
The data of the discontinuities found on the RPV
surface were entered in a data basis and systematized.
Following 15-17 years from the first start-up of a reactor
unit, the first discontinuities in the metal structure (that
may actually be detected by inspection methods) can
be observed on the inner surface of the reactor pressure
vessel. With the progress of the unit’s operating course,
local merges of discontinuities can be observed.
Corrosion-erosion foci are formed, concentrated in the
fretting area of the strengthening nodes of the reactor
internals, as is shown on “Figure 8”. Single surface
defects are observable; the defect parameters get
determined.
Figure 8: Discontinuities on the Inner Surface of the Reactor Pressure Vessel
The corrosion rate is one of the ageing effect indicators.
The corrosion rate monitored indicators were:
• Beginning of the occurrence of corrosion outbreaks;
The data on this indicator are decisive for the start
of monitoring of the area.
• Changes in the size of the discontinuities monitored.
• Occurrence of new defects.
The periodicity of this inspection is specified in
the technical specifications for operation of the nuclear
power unit. Usually, it is once in 4 years. However, each
NPP may alter it at its own discretion.
The respective parameters are identified for
each defect and each group of defects. Each
assessment of the parameters is followed by an analysis
of the acceptability of the defects in terms of the
normative requirements [23, 24]. The analytical part of
the activity can offer recommendations of the future
operation of the reactor equipment. Normally, these
recommendations regard, as follows:
• Mechanical activities - the reactor internals shall be
positioned in a way so as to prevent mechanical
scratching, scuffing, denting, etc.
• The water chemistry shall be more benign to the
surfaces.
c) Study of RPV Metal Embrittlement Resulting from
Occurrence of a Discontinuity
The metal on the inside of the reactor pressure
vessel shall be tested (controlled) once every four years
(as required by normative regulations). In case defects
are present or indications of defects, it is recommended
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Ageing of Nuclear Power Plant’s Equipment. Assessment of Reactor Pressure Vessel Ageing Effect

that this interval be shorter. Visual inspection, penetrant
testing and UT are the control methods usually applied
by means of remotely controlled scanning technical
tools. Non-destructive examination methods enable
finding any discontinuities (defects, cracks) and
studying their parameters, i.e., location, type, size,
orientation.
At the places where discontinuities have been
found studies are performed to identify the
environmental load factors, what their values are, and
whether they change with the varying design operational
modes of the reactor unit. The operating conditions (for
the inner surface metal) are characterized by intensive
neutron flux with neutron energy exceeding 1.5 MeV;
high pressure values (up to 17.5 MPa; and high
temperatures of the primary circuit fluid (323oC).
Evaluations are conducted on the impact of the
loads on the evolution of the defects. The discontinuities
are studied by the visual inspection and ultrasonic
testing, while the studied parameters are location, type
and size (area). The input data of this study cover data
of the defects and of the loads in effect:
• The defects indications data include: location, type,
size (,), distance  from the internal surface of
the reactor pressure vessel. The data for the defects
and defect indications result from applying the non-
destructive examination methods.
• Data of the neutron fluence are collected on an
annual basis by monitoring the readings of the
neutron detectors (reactor internal ones). Another
data source are surveillance specimens that are
withdrawn from the RPV and tested according to a
dedicated program.
• Data of the active stresses can be obtained from the
RPV datasheet and/or strength analyses of the
manufacturer (conducted on unit “a”).
• The load factors in the zone where defects occurred
can be obtained, as follows:
• From the readings of neutron fluence detectors for
the inner surface of the RPV;
• The active stresses values are obtained from the
strength analyses (calculations) and equipment
datasheets.
A graphical presentation method is
implemented to demonstrate the neutron fluence
distribution on the inner surface of the RPV metal. A
coordinate system is used. Along the “x” axis the
coordinates of the RPV inner surface are marked (х=0
indicates a location on the innermost layers of the
reactor); the fluence values are shown on the “y” axis.
The graphic representation method visualizes the
fluence distribution along the thickness of the reactor
wall.
Using a graphic method, the distribution of
circular stresses is shown as depending on the distance
“X” from the border of the deposit weld metal with the
base metal of the RPV inner surface. Similarly, a graphic
method is used to demonstrate the distribution of
thermal stresses as dependent on the distance X from
the border of the deposit weld metal with the base
metal.
Selection of defects means the parameters of
all the identified discontinuities are reviewed. An
assessment is made to decide which of them are
located in zones with a degradation potential. Large size
defects present particular danger. The discontinuities
located close to the inner surface of the RPV are
considered to be subject to the comprehensive impact
of the working environment loads inside the pressure
vessel, i.e., high values for the neutron fluence and
thermohydraulic loads. The discontinuities that have an
opening to the RPV inner surface present a hazard for
occurrence of intracrystalline (intergranular) corrosion,
stress corrosion, etc. Of great significance is the
orientation of the discontinuity with regard to the
direction of the active loads; the most dangerous are the
type A stresses (crack resistance).
Several of the “most dangerous” discontinuities
are selected and calculations are made about them. For
each of the selected discontinuity calculations are made
to obtain: 1) the values of the stress intensity factor, and
2) the critical values of the stress intensity factor.
Brittle fracture toughness is ensured if the following
condition is met with regard to the discontinuity found
[19]: [] (6)
=.. (7)
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where stands for the load, and is a coefficient
related to the discontinuity shape, аis the small semi-
axis of the discontinuity, as is shown on the “Figure 9”.

Figure 9: A diagram of a below-surface discontinuity; a - small semi-axis.
For a below-surface discontinuity in point A, of “Figure 9”:
=
1.79 0.66.󰇡
󰇢
[1 1.8 . (1 0.4. 
0.8. 0.4 )]0.54
=
+ ; = 0.5 +
; +
2
=3+
4+
().󰇡
󰇢
12
For a below-surface discontinuity in point B, of “Figure 9”:
=
1.79 0.66.󰇡
󰇢
[1 1.8 . (1 0.4. 
2)]0.54
=+ 3. 
4+
().󰇡
󰇢
12
For a surface discontinuity in point A, of “Figure 9”:
Y= 20.82.󰇡
󰇢
1󰇧0.890.57 .
󰇨3
.󰇡
󰇢1.53 .25
= 0.61.+ 0.39.+0.11.
0.28.
.1
.()
For a surface discontinuity in point B, of “Figure 9”:
Y=󰇣20.82.󰇡
󰇢󰇤.󰇩1.1+0.35 .󰇡
󰇢2.
󰇪
1󰇧0.890.57 .
󰇨3
.󰇡
󰇢1.53 .25
= 0.18.+ 0.82.
For welded joints:
[]=25 +27.exp 0.0235. () in hydraulic tests mode (8)
[]=35 +53.exp 0.0217. () in accident conditions (9)
S
2cb 2a
AB
C
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The study was conducted over the years 1993 -
2014. Remotely operated equipment for visual
inspection and ultrasonic testing scanning type of
equipment that permits sequential sounding of all parts
of the reactor pressure vessel (object of control). The
sounding (a signal is input in the metal and then the
reflected signal is registered) is remotely operated.
Software instrumentation is employed to register the
results. All identified images for defect indications are
stored in the memory, their size is taken as well as
coordinates. To complete this activity a UT system, type
P-scan, for the RPV inner surface is used together with a
type Tomoscan system.
The results are obtained using the following algorithm:
1) Regular, periodic non-destructive examination of the
reactor pressure vessel is conducted implementing
inspection methods, i.e., remote visual inspection
and ultrasonic testing). This NDE is held once every
4 (four) years.
2) If indications of discontinuities have been found,
their parameters shall be identified, i.e., length,
location and size of equivalent area (UT
characteristics of the indications).
3) If discontinuities have already been found during
preceding NDE campaigns, the discontinuity
indications’ parameters are measured again.
4) A data base is established and the indications’ data
are input in it following each NDE of the RPV.
5) The data of the discontinuity indications are
arranged by 1) size of the length, 2) area, 3)
location, 4) the RPV operating period (in years). The
location of the indications in the RPV metal is
registered in a 3-D coordinate system - height, RPV
length and depth of embedding in the metal, as
read from the RPV inner surface.
6) A screening of the discontinuities is performed for
the purpose of further assessments and
calculations. The indications having the largest area
have their location tracked in terms of the distance
from the RPV inner surface. The location is an
important factor as the values of the fluence and the
thermohydraulic loads tend to change in the
different points of the RPV.
7) A graphic method is implemented to demonstrate
the neutron fluence distribution on the inner surface
of the RPV metal.
8) As regards the fixed locations of the indications
(critical zones), the circular stresses and
temperatures are considered under the different
operating modes of the reactor unit. To ensure
conservatism of the calculations, the highest values
are used for: 1) circular stresses, 2) temperatures
under all the design modes.
9) The number of years elapsed since the start of
operation of the pressure vessel are taken into
account. The operating period is important for the
evolution of the discontinuities - whether the thermal
embrittlement or the neutron one have greater
impact, insofar as these influences can be
considered separately in the analyses.
10) Calculations are made of the stress intensity factors
 for selected indications. The critical temperature
shift (,) is calculated using two methods: 1)
according to the strength norms [19] and 2)
according to the European documents [11].
11) Calculations are made for the critical values of the
stress intensity factors .
12) The results obtained for the stresses intensity
factors  are compared with the critical ones [],
as in “Equation (6)”.
A graph has been prepared demonstrating the
distribution of the relative values of the neutron fluence
along the depth of the RPV, as shown in “Figure 10”.
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Neutron fluence
F/Fmax
Distance
X, [mm]
1
0,9
0,8
0,7
0,6
0,5
0,4
0,3
0,2
0,1
20 40 60 80 100 120 140 160 180
Figure 10: Distribution of the Relative Values of the Neutron Fluence F/Fmax on the Inner Metal Surface of the Reactor
Pressure Vessel
Non-destructive visual inspection and ultrasonic
testing were performed. The indications parameters,
such as size, location coordinates, year of size taking,
were assessed. The data of a welded joint indications
and the relative values of the neutron fluence F/Fmax are
shown in “Table 2”.
Table 2: Indications of Welded joint 2 of the Reactor Pressure Vessel – Sizes, Coordinates and Neutron Flux at the
Position ot the Indicates
Size and coordinates: a) Along the weld.
b) Along the reactor height.
c) In depth of the weld, on the outside.
Indication №
1 2 3 4 5 6
Size [mm] 478 39 39 150 55 27
a) [grad] 64.4 79.9 85.4 88.4 104.9 107.4
b) [mm] 274.5 276.3 277 276.7 276.8 277
520 52 53.7 51.4 51.6 50.6
Neutron flux F /
Fmax 0.15 0.15 0.15 0.15 0.15 0.15
Indications were identified at two more welded
joints of the reactor pressure vessel. To forecast the
development of the discontinuity’s indications in the
metal, the typical transitional state of “Accident
conditions - Primary circuit large leak” has been
considered. Pursuant to the register of the implemented
operational cycles, the large leak mode is associated
with the highest amplitude values for stress-temperature
fields.
“Figure 11”
demonstrates the relative
distribution of circular stresses as dependent on the
distance X from the border of the deposit weld metal
with the base metal.
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Circular
stresses,
[MPa]
X,
[mm]
400
300
200
100
0
-100
-200
-300
20 40 60 80 100 120 140 160 180
Х
Х
Х
Х
Х
Х
Х
Х
Х
Х
Х Х
2
О
О
О
О
О
О
О
О
О
О
ОО
3
1
Figure 11: Relative Distribution of Circular Stresses as Dependent on the Distance X from the Border of the Deposit
Weld Metal with the Base Metal. X=0 is the Position that is Closest to the Inner Surface of the RPV; the Active
Stresses have Maximum values. Curve 1 – at the Moment of 0.2 hrs of the Time Interval of the Mode; Curve 2 – at 0.4
hrs; Curve 3 – at 0.6 hrs
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“Figure 12” shows the temperature distribution in the RPV metal depending on the distance “X” from the
border of the deposit weld metal with the base metal.

Temperature, ⁰ C
X, [mm]
300
250
200
150
100
50
20 40 60 80 100 120 140 160 180
1
Х
Х
Х
Х
Х
Х
Х
Х
Х
Х2
О
О
О
ОО
О
О
ООО3
Х
Х
О
О
Figure 12: Temperature distribution in the pressure vessel metal as dependent on the distance X from the border of
the deposit weld metal with the base metal. Primary circuit large leak mode: Curve 1 – at the moment 0.2 hrs of the
time interval of the mode; curve 2 – at 0.4 hrs; curve 3 – at 0.6 hrs
Table 3: Stresses Values at the Locations of Discontinuities σϴBH - Circular Stresses of the Base Metal Inner Surface
Weld joint/
Indication
№
Distribution of circular stresses at the locations of discontinuities
for the “Primary circuit large leak” mode, σϴBH [МРа]
0,2 hrs 0.4 hrs 0.6 hrs
2 / I -210 -170 -160
2 / II -210 -170 -160
2 / III -210 -170 -160
2 / IV -210 -170 -160
2 / V -200 -175 -160
2 / VI -240 -190 -180
3 / I 487 350 260
3/ II -80 -175 -150
3 / III -110 10 -35
3 / IV -250 -190 -185
4 / I 450 140 130
4 / II 210 250 200
4 / III -200 -115 -120
Table 4: Temperature Values at the Locations of Discontinuities.
Weld joint/
Indication №
Distribution of temperature in the RPV metal for the “primary
circuit large leak” mode, [°C]
0,2 hrs 0.4 hrs 0.6 hrs
2 / I 260 180 158
2 / II 260 180 158
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The values of the circular stresses and the temperature at the locations of discontinuities are provided in
“Ta b l e 3 ” and “Table 4”.

Weld joint/
Indication №
Distribution of temperature in the RPV metal for the “primary
circuit large leak” mode, [°C]
0,2 hrs 0.4 hrs 0.6 hrs
2 / III 260 180 158
2 / IV 260 180 158
2 / V 260 178 150
2 / VI 250 190 160
3 / I 70
55 45
3/ II 240 177 145
3 / III 210 160 130
3 / IV 250 190 167
4 / I 150
110 70
4 / II 108 65 60
4 / III 240 170 140
For the purpose of this study, calculations were
made of the stress intensity factors under accident
conditions and “primary circuit large leak” mode. The
stress intensity factor, a below-surface discontinuity is
calculated with the “Equation (7)”.
A screening was performed to select the
discontinuities have to represent the worst-case
scenario. Two indications were chosen of weld joint  2
that are largest in size compared with the rest of
discontinuities identified, and are located in areas where
high temperature of the metal is expected in the primary
circuit large leak mode. Also, an indication of welded
joint  3 was singled out as it was large in size and was
located in close proximity to the inner surface of the
RPV. The stress intensity factors K were calculated for
selected indications. The results for the calculations of
are provided in
“Ta b l e 5 ” .
Table 5: The Results for the Calculations of  Under Emergency Condition, Primary Circuit Large Leak Mode
Weld joint/
Indication № 
in point A, circular stress, under emergency condition, primary
circuit large leak mode, [MPA.]
0,2 hrs 0.4 hrs 0.6 hrs
2 / I 21.02 17.02 16.02
2 / IV 23.8 19.2 18.1
3 / I 60.21 43.27 32.14
Calculations were made of the limit values of
the stress intensity factors [] under emergency
condition, primary circuit large leak mode. “Table 6
”
Table 6: The Results for the Calculations of  and [], under Emergency Condition, Primary Circuit Large Leak
Mode
Weld joint/
Indication №

in point A, circular stress, under emergency condition, primary
circuit large leak mode, [MPA.]
[] [] per
[19] [] per [11, 25]
3 / I 60.21 108 123
For Indication №1 of welded joint 3 a
comparison was made between the current values of
the stress factor and the limit value[]:
={60,21; 43,27; 32,14}<[] = {108,123}
Meeting the condition of “Equation (6)” is a
regulatory criterion that fracture of the weld joint will not
occur as a result of brittle fracture due to the present
discontinuity in case of accident conditions, with primary
circuit large leak mode.
VI. Conclusions
Currently, metal ageing issues have been
gaining increasing topicality as the equipment in most of
the operating Nuclear Power Plants is already “aged”,
and although a good deal of knowledge has been
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contains the results about indication №1 of welded joint
№3.

130.
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Ageing of Nuclear Power Plant’s Equipment. Assessment of Reactor Pressure Vessel Ageing Effect
accumulated on NPP components ageing, it is
becoming clear that there are still a great many
unexplored areas.
NPP equipment and components are subject to
the continued impact of stress factors. The mechanisms
of the mechanical properties’ degradation of WWER
reactor type metal components are corrosion, erosion,
neutron and thermal ageing, fatigue and wear. These
impacts induct changes in the mechanical properties of
metal and, eventually, result in the component’s loss of
operability. The latter could compromise the NPP safety
and also cause economic losses due to generation
losses. The degradation mechanisms of mechanical
properties have been individually researched into a
laboratory conditions, while materials in NPP are
subjected to integrated impact of load factors. In
general, the proposed lifetime characteristics
assessment methodologies are based on the
independent occurrence of processes such as
corrosion, fatigue, creep and neutron embrittlement,
although in reality these processes run simultaneously in
various combinations [26].
The paper proposes a methodology for study of
ageing effects. A case is presented of implementation of
the methodology for assessment of ageing effects of
reactor pressure vessels. The RPV is subject to the
impact of multiple load factors.
Of all the degradation mechanisms that affect
the RPV, the radiation and neutron fluence thermal
impacts are the most destructive ones. Over the first
decade of an NPP operation, thermal embrittlement is
dominant, followed by neutron embrittlement. The nickel
(Ni) content enhances the embrittlement trend which is
manifested in the growing value of the embrittlement
temperature shift. Greater embrittlement occurs in weld
metal than in base metal.
The evaluation of the ageing effects points out
that one of the greatest hazards for the integrity of the
reactor pressure vessel is the presence of a
defect/defects on account of them posing the possibility
for brittle fracture. The following cases of RPV metal
defects were studied: 1) surface defects of base metal;
2) defects of weld joints in the maximum neutron fluence
areas (dominant factor); 3) defects in areas with high
values of stress and temperature (dominant factor).
Based on the studies performed it can be
concluded that case 2) poses the greatest hazard.
Defects in metal on account of equipment
ageing occur following a long term of operation. Before
extending the NPP lifetime it is required to confirm the
operability of components. Equipment condition
investigation is a challenge since, clearly, one cannot
destroy non-replaceable equipment of an NPP in
operation in order to examine its degree of degradation.
Therefore, it is important to track the manifestation of
ageing effects throughout all stages of NPP operation.
Acknowledgment
The scientific research, the results of which are
presented in this publication, were financed by the
internal competition of TU-Sofia-2023
References Références Referencias
1. Krivanek, R., Fiedler, J. (2017) Main corrective
measures in an early phase of nuclear power plants
preparation for safe long term operation. Nuclear
Engineering and Design Journal, 316, 125-
https://www.sciencedirect.com/journal/nuclear-eng
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