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Copyright © 2007
AREVA NP & EDF
All Rights Reserved
This document is the property of AREVA NP and EDF. It contains
proprietary information and, except for affiliate companies of AREVA NP
and/or EDF, may not be reproduced or copied in whole or in part nor may it
be furnished to others without the prior written permission of AREVA NP and
EDF. In any case, any re-exportation of the document shall be subject to
the prior written permission of AREVA NP and EDF. It may not be used in
any way that is or may be injurious to AREVA NP or EDF. This document
and any copies that may have been made must be returned on request.
The design, engineering and other information contained in this document
have been prepared by or on behalf of AREVA NP and EDF in connection
with their request for generic design assessment of the EPRTM design by the
UK nuclear regulatory authorities. No use of or right to copy any of this
information, other than by the UK nuclear regulatory authorities and their
contractors in support of generic design assessment of AREVA NP’s and
EDF’s joint submittal, is authorized.
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For information address:
EDF
AREVA NP SAS
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Tour AREVA
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1.
GENERAL REMARKS
This Design and Safety report forms Volume 2 of the submission produced by EDF and Areva
as part of the Generic Design Assessment process for the UK EPR. The material presented in
this Volume is extracted from the public version of the preliminary safety report (PSAR)
submitted to the French Public Authorities in support of the application to construct the
Flamanville-3 (FA3) EPR unit in France. Information specific to the FA3 site that would not be
applicable or relevant to a UK sited EPR has been removed.
This chapter provides the definitions of the terms employed and the abbreviations used in the
Volume 2 Design and Safety Report. To maintain consistency with the FA3 EPR preliminary
safety report, French abbreviations are retained throughout Volume 2, but where applicable, the
equivalent English abbreviation is also given, identified by a square bracket. Where the English
abbreviation is the same as the French, or where there is no English abbreviation in common
use, the English abbreviation is omitted.
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2.
DEFINITIONS
Activity (Radioactive)The activity is the number of nuclear disintegrations in a radionuclide per unit of time
In the SI system, the unit of activity of a radiation source is the becquerel (Bq), which is equal to
the activity of a quantity of radionuclide for which the mean number of nuclear disintegrations
per second is 1.
1 Bq = 1 s-1
Conversion of the activity values into the non-SI unit, the curie (Ci), is such that:
1 Bq = 2,7027 x 10-11 Ci
1 Ci = 3,7 x 1010 Bq
Activity, specific
The activity of a material per unit mass.
Activity concentration
The activity concentration is the activity per unit of volume of a material or a fluid (liquid or
gaseous) in a system or a portion of a system
ALARA (As Low As Reasonably Achievable)
An approach whereby the protective measures against ionising radiation are designed and
implemented so that exposure of personnel, in terms of probability and magnitude, are at the
lowest level that can be reasonably attained, taking into account economic and social factors.
This approach is required by the CIPR [ICRP] (Commission Internationale de Protection
Radiologique=International Commission for Radiological Protection).
Alarm
An alarm is an alert signal produced by the instrumentation and control systems which is
retransmitted to the control room to notify the operators of the occurrence of a operational fault
or a change in the plant state which requires one or more actions by the operating team.
Area dose rate
The intensity of radiation in vicinity of detectors that is considered representative of the general
level.
Area dose rate monitoring
This is performed by the local measuring equipment including the detection head (for example
scintillation counter) and its signal transmitter.
Barrier
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Any device placed between radioactive substances and the environment in order to prevent or
restrict dispersal.
Guillotine Break
A guillotine break refers to a break involving complete severance of a pipe.
Break (Rupture) Preclusion
Break (Rupture) Preclusion is a deterministic concept, which assures by preventative measures
(requirements on design, materials, product forms, manufacturing, quality assurance, analysis
of fatigue crack growth) and surveillance measures (requirements on transient and water
chemistry monitoring, leakage detection, in-service inspection) that rupture of a pipe can be
ruled out.
Burnup ratio
The ratio, usually expressed as a percentage, of the number of nuclei of one or more chemical
elements which disappear through burnup in relation to the initial number of nuclei.
Busbar, bus
Busbar, or bus, is a generic term designating connection devices or switchboards used to
distribute a voltage level.
Civil Steel Structure
Set of steel structures which do not act as an equipment support but rather are used for access
to equipment, for example platforms, columns and stairwells.
Common cause failure
The failure of at least two structures, systems or components in the execution of their function
due to a single specific event or cause
Component
A component is a clearly defined part of a system capable of performing specific sub-functions.
Examples of mechanical components are tanks, heat exchangers, pipes, pumps and valves.
Component, active
An active component is a component operated or controlled externally and activated manually
or automatically with the assistance of transfer and driving media (e.g. electric current, or
hydraulic or pneumatic actuation system). A self-acting component (i.e. one that operates
without an external power or control) is considered to be an active component if its position
changes when fulfilling its intended function (e.g. a pressure relief valve).
Components, high-energy
High-energy components are those water-, steam carrying components the pressure of which
during normal operating conditions of the plant is equal or greater than 20 bar or the
temperature of which during normal operating conditions of the plant is equal or greater than
100°C. Gas carrying components with a pressure beyond atmospheric pressure are always
considered as High-Energy Components. All other components are referred to as low-energy.
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Component, passive
Passive components need no actuation and energy supply to perform their function (e.g. pipes,
heat exchangers, capacities, simple non-return valve).
Construction work
Everything that is constructed or results from construction operations. It covers civil engineering
work and other engineering works. It refers to the complete construction including both
structural and non-structural elements.
Controlled area
Area in which workers are likely to receive an annual dose higher than 6 mSv under normal
working conditions Controlled areas are divided into four zones: green, yellow, orange and red,
according to the relevant dose rate.
Controlled state
For type PCC-2 to PCC-4 events, the controlled state is defined as a state when the fast
transient is finished and the plant is stabilised. This corresponds to:
−
the core has been made sub-critical (short term re-criticality, before operator actions
with only low power can be accepted , on a case by case basis),
−
Decay heat is being removed (for example by steam generators),
−
the core cooling water inventory is stable,
−
activity releases remain at an acceptable level.
For PCC-2 to PCC-4 type events affecting the spent-fuel pool, the controlled state is defined as
a state characterised by the establishment of decay heat removal from fuel stored in the pool.
Core damage, conditional frequency
The conditional frequency of core damage is conditional upon certain initiating event. The
frequency is calculated by rerunning the probabilistic safety analysis model after setting the
unavailability of basic events associated with the inoperable components to “True”, and also
adjusting the relevant common cause failure parameters, to take account of equipment
unavailability due to the initiating event.
Core meltdown, conditional probability
The conditional probability of core meltdown is conditional upon some event. This probability is
calculated by rerunning the probabilistic safety analysis code after setting the unavailability of
basic events associated with inoperable equipment components equal to “True”, and configuring
the appropriate common cause failure parameters, to take account of equipment inoperability
due to the initiating event.
Core meltdown, probability
The probability of core meltdown within a time interval is obtained by multiplying the frequency
of core meltdown by the length of the time interval.
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Corium
Corium refers to the agglomeration of materials produced by core meltdown. Its detailed
composition will depend on the core meltdown scenario, but it will normally contain UO2, fuel
element cladding material (Zr and ZrO2), materials of absorber rods (Cd, B, Ag and steel), steel
from core components and from the pressure vessel, and fission products.
Crack, stable
A stable (or sub-critical) crack is a longitudinal or circumferential crack whose size does not
increase under a given stress or only propagates slowly when it is subjected to a fluctuating
stress. The length of a stable crack at the specified stress is less than the critical crack’s length.
Crack, critical size
The critical crack size is the length of the crack beyond which spontaneous crack propagation
occurs (at a high propagation rate). The critical crack size depends for example on the stress
applied, the material properties and the operating temperatures, and is determined with the aid
of fracture mechanics.
2% criterion
The 2% criterion is a criterion which allows to rule out breaks. It applies to qualified piping
systems of more than 50mm nominal diameter which are in operations as high energy systems,
only for period of demands less than or equal to 2% of the plant lifetime.
Critical Heat Flux ratio
Within the reactor core, the critical heat flux ratio is the ratio between the critical heat flux and
the actual heat flux at a given point of the surface of the fuel cladding.
Criticality
A medium containing a fissionable nuclear material is critical if a nuclear chain reaction is
automatically maintained, i.e. if the number of neutrons is stable (production rate =
disappearance rate by absorption and external leakages).
Crust formation
An oxidation crust is formed on the surface of the corium when it cools down. Owing to the low
heat conductivity of the crust, the subsequent cooling process can be retarded or even
terminated.
Design
The design of a component (or of an elementary part) includes the definition of the function or
process performed by the component, the role played by the component in performing the
function, the layout, the choice of materials and the design of the component.
Direct containment heating [DCH]
The phenomenon of the direct transfer of thermal energy from corium into the containment
atmosphere by dispersing the corium droplets or aerosol during corium discharge from the
pressure vessel.
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Diversity
Diversity is the existence of redundant systems, structures or components capable of
performing an identified function, which have different attributes so as to reduce common cause
failure (including common mode failure) following an initiating event.
Division
This term concerns the layout or the configuration of buildings. The divisions are segregated so
that the systems within each division are separated by distance or a barrier and thus maintain
physical, electrical and functional independence from other redundant systems performing the
same function in the other divisions.
Dose / absorbed dose
The dose/absorbed dose is the energy imparted to matter by ionising radiation when it passes
through an inert material. In the SI system, the unit of absorbed dose is the gray (Gy), the
absorbed dose in a 1-kilogram mass of matter to which ionising radiation imparts uniformly on
average an energy of 1 Joule:
1 Gy = 1 J/kg
As a reminder, the conversion of values of absorbed dose to rad, which is the old unit, is as
follows:
1 rad = 10-2 Gy
1 Gy = 100 rad
Dose / Dosimetric impact
The dosimetric impact is the total effective dose received during a 50 years period by a whole
organism (adult or child) due to 4 types of exposure: radioactive plume, inhalation, exposure to
deposited activity and ingestion.
Dose / Effective dose
The effective dose takes into account the different sensitivity of various organs and tissues to
ionising radiation. The effective dose is the sum of the equivalent doses transmitted to the
various organs and tissues weighted by factors specific to each organ or tissue.
The unit of effective dose is the Sievert (Sv).
As a reminder, the conversion of the equivalent dose to rad, which is the old unit, is as follows:
1 Sv = 1 J/kg = 100 rem
Dose / Equivalent dose in an organ or tissue
The equivalent dose is the product of the dose absorbed in an organ or tissue by a weighting
factor which depends on the type of radiation.
In the SI system, the unit of equivalent dose is the Sievert (Sv), which corresponds to an
absorbed dose of 1 J/kg for photons.
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As a reminder, the conversion of the equivalent dose to rad, which is the old unit, is as follows:
1 Sv = 1 J/kg = 100 rem
Earthquake design (SDD)
This term refers to the design of the civil engineering structures or equipment against seismic
events. Seismic parameters (ground spectrum, level and characteristics) are specified under the
generic term design spectrum, which specifies each of the components of the horizontal
acceleration in the design basis earthquake. The installation must be designed against loads
enveloping those caused by movements associated with the safe shutdown earthquake (SMS)
[SSE]. (See safe shutdown earthquake)
Electrical Power division
The electrical power division is an independent functional part of the normal electrical
distribution network.
Electrical supply train
An electricity supply train includes all the components required for the distribution of electricity.
Electricity supply trains are independent, and are each allocated to a section within the
conventional island or a division within the nuclear island.
Embedded steel parts
This term refers to steel components embedded within a reinforced concrete wall which ensure
the interface between the supports and the construction works. Examples are anchor plates,
studs and bolts.
Emergency power supply mode
Operating state of the power plant during which electrical equipment is supplied by the
emergency diesel generators only.
Equipment
The term equipment designates all the devices, mechanisms, means or resources (except staff)
enabling a system to fulfil its function.
Ergonomics
Ergonomics (or the study of Human Factors) is a scientific discipline dealing with the
interactions between humans and system components, and applying it to design theories,
principles, methods and relevant data in order to improve the welfare of man and the global
effectiveness of systems.
Exclusion
The exclusion of a particular event is the result of a demonstration showing that such an event
can be ruled out from the general scope of events addressed in the design basis of the unit. In
practice, exclusion mainly applies to accident sequences that could lead to large early releases.
(See Practical elimination).
Ex-vessel cooling
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This refers to stabilising / cooling of a molten pool of corium outside the reactor pressure vessel.
Fault
A fault is a defect in hardware, software or system component. Random faults can be caused by
hardware wearing out. Systematic faults may occur due to design errors or to initial equipment
imperfections etc. In the case of software, systematic faults may occur due to coding and
specification errors.
Final state for RCC-A Analysis
For RRC-A accidents involving multiple failures, the final state is defined as:
−
−
−
core sub-critical,
decay heat removed by primary or secondary systems,
activity release maintained at an acceptable level.
Fire barrier
Fire barriers are fire resistant structures and components such as walls, floors, ceilings, ducts
and closures such as doors, smoke dampers, fire dampers, air locks and penetration sealing of
the cable trays and pipes.
Fire zone
A fire zone is a subdivision of a fire compartment, delimited by partitions or boundaries that
guarantee that a fire arising inside cannot spread outside or that one arising outside cannot
spread inside.
Fire compartment
A fire compartment is a building or part of a building comprising one or more rooms, delimited
by partitions whose fire resistance guarantees that a fire arising inside cannot spread outside or
that one arising outside cannot spread inside. All the partitions of a fire compartment must be
fire resistant.
Fire volume
One of the ways to prevent the spread of fire is to keep it in a confined space, either by using
physical partitions such as fire compartments, or by defining virtual boundaries based on the
distance between components, and by using active protection systems (sprinklers) and passive
protection systems (structural elements and enclosures), to delimit a fire zone.
Fragmentation
This term designates the disintegration of the corium into small particles. This phenomenon can
be achieved by corium falling into water, or by water, steam or gas injection into the corium.
Functional capability
The ability of all the pressure-bearing parts of a component (active or passive), to withstand the
specified loading without experiencing a deformation large enough to impair its safety function
or reduce its capacity.
Generator circuit breaker
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The circuit breaker located between the step-down transformers (on the high or low voltage side
of the main transformer) used to isolate the generator from the station electrical network, for
example on normal unit start-up.
Grid breaker
The grid breaker is a circuit breaker located between the power plant and the external grid and
is used, for example. to isolate the plant from the grid during house load operation.
Grid connection, Alternative
This refers to the connection of the plant to the external grid in order to receive the required
energy for plant shutdown, in the event of the loss of main grid connection containing all
elements between auxiliary lines and low voltage terminal of auxiliary [station] transformer.
Grid connection, main
This refers to the connection of the plant to the external high voltage (HTB >50kV) system in
order to transmit electrical energy produced by the plant on to the grid network. The connections
include all elements between the HTB terminals of the main transformer and the HTB terminals
of the step-down [unit] transformers.
H2 deflagration
Deflagrations are combustion waves in which the unburned gases are heated by thermal
conduction across the wave-front to temperatures high enough to produce a chemical reaction.
Deflagration waves normally propagate at subsonic speeds and result in quasi-static (nearly
steady state) loads being applied to the containment.
H2 detonation
Detonations are combustion waves in which heating of the unburned gases is caused by shock
wave compression. Detonation waves travel at supersonic speed and produce impulsive or
dynamic loads on the containment which are additional to quasi-static loads.
H2 Recombiner
A piece of equipment in which hydrogen reacts with oxygen below the self-ignition temperature
in the presence of a catalyst, to form water.
Hazards
Internal hazards are events that originate on the plant site which have the potential to cause
adverse conditions or damage on the inside of safety classified buildings (e.g. internal fires or
flooding, internal missiles or explosions, etc.). Such effects can result in common cause failures
within systems used to reach or maintain the unit in a safe state.
External hazards are natural or man induced events originating outside the plant with the
potential of adversely affecting the safety of the plant, and of leading to radiological
consequences. Such hazards include earthquakes, aircraft crash, external explosions and
external flooding etc.
Heat sink
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The atmosphere, a body of water, or a combination of both, into which residual heat is
transferred.
Heat sink, Main Removal
The mechanical system necessary for keeping the steam generators and the main condenser in
operation to evacuate heat produced by the reactor (e.g. circulating water, start-up and
shutdown feed pumps and condensate extraction pump).
Heat sink, ultimate
Heat sink to which heat is transferred by the Essential Service Water System [ESWS].
Heat sink (Ultimate Cooling Water System) [UCWS], Reserve Ultimate
Heat sink into which residual heat can be transferred in the event of unavailability of the ultimate
heat sink.
High-pressure core meltdown / High Pressure Path
A type of core meltdown scenario which causes high pressure failure of the primary circuit.
House load operation
Mode of operation of the plant, following a grid problem, where the electrical energy produced is
used only to supply its own electrical auxiliaries, without a link to the grid.
Human Factors Engineering [HFE]
A term used by analogy with the term engineering, to designate the consideration of Human
Factors into applied design engineering. This terminology covers objectives, a method, a field of
application and a programme. A Human Factors Integration Programme is also referred to.
Human-machine interface [HMI]
The methods by which the operator interacts with a process. The human-machine interface
includes the use of conventional or computerised methods for the control and monitoring of the
installation (indicators, recordings, alarm windows, displays, controls, etc.)
Ionizing radiation
Radiation composed of photons or particles which are capable of causing the formation of ions
either directly or indirectly when passing through matter.
Independence
A system or equipment is independent if it has both the following characteristics:
−
−
the ability to perform its required function is not affected by the operation or
failure of other equipment;
The ability to perform its function is not affected by the presence or the
effects resulting from the postulated initiating event for which it is required to
function;
Initiating event (or postulated initiating event [PIE])
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An event potentially affecting the power, the reactivity control, reactor heat removal or
containment of radioactive material (PCC-2 to PCC-4 or RRC).
Inspection earthquake (level of)
After the occurrence of an earthquake up to this level, no verification and inspection of the
components important to safety would be needed prior to return or to maintain the plant to
normal operation. Adequate provisions must be made at the design stage to allow appropriate
inspection and testing to be carried out should an earthquake occur which exceeds the
Inspection Earthquake level.
Integrity
Integrity is the ability of all elementary pressure-bearing parts of a component to withstand the
specified loadings at a given frequency of occurrence during the lifetime of the component.
Late water injection
The injection of water in a core meltdown accident after the core is already severely damaged.
Leakage
Leakage results from a total localised loss of wall thickness in a pressurized component, either
due to a stable (sub-critical) wall crack, or to the loss of wall thickness (erosion-corrosion). A
leak results in the loss of containment capability for the pressurised component without
occurrence of a break in the pressure boundary.
Leak before break [LBB]
Leak before break describes the situation in which a leak occurs before a complete doubleended break of a component. Although not considered by itself as sufficient to demonstrate
rupture preclusion, which is dependent on the component integrity, it can provide one element
of the safety argument to limit the potential consequences associated with the loss of integrity
and provide information on the strength of the pipework.
Limitation
An automatic function of a preventative nature intended to introduce gradual corrective actions
to avoid actuating a protection system, thereby improving plant availability.
N.B.: Limitations are not taken into account in carrying out deterministic safety analyses. (See
LCO)
LCO [Limiting Condition of Operation]
The LCO concept combines several ideas, namely:
−
−
−
the LCO bounding value which corresponds to the value assumed in accident
analyses,
LCO monitoring function, which involves controlling parameters within their normal
operating range (around their set points),
LCO function, which designates the control command functions specifically
implemented to initiate (manually or automatically) counter-measures in the event of
a parameter departing from its normal operating range.
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Liquefaction
Liquefaction is a sudden loss of ground rigidity and of ground resistance to shearing when an
earthquake occurs, due to ground saturation and loss of cohesion.
Loss of primary coolant accident (APRP) [LOCA]
APRP [LOCA] is characterised by a leak or a break in a primary system pipe or by the
inadvertent opening of a pressure relief valve or of an isolation valve which results in a loss of
primary coolant beyond the capability of the primary circuit water makeup system. APRP
[LOCA] covers break sizes as defined below:
-
Small break [SBLOCA]: a break with an equivalent diameter less than or equal to
50 mm;
-
Intermediate break [IBLOCA]: a break with an equivalent diameter greater than 50
mm (equivalent cross-sectional area greater than 20 cm2), which is smaller than a
large APRP [LOCA] break;
-
Large break [LBLOCA]: a “guillotine” type break of the largest pipe connected to an
RCP [RCS] Loop. This corresponds to a surge line break on the hot leg side or an
RIS [SIS] line break on the cold leg side.
-
APRP 2A [2A LOCA]: a double ended break of a reactor coolant system Loop pipe:
‘A’ designates the cross-sectional area of the piping.
Main secondary system (CSP) [MSS]
The main secondary system of the nuclear steam-supply system comprises the secondary
volumes within the steam generators and the all the pipework which cannot be safely isolated
from them, including safety equipment and pressurized equipment which performs an isolating
role.
Maintenance
Maintenance includes all the actions, technical, administrative and managerial carried out during
the life cycle of an equipment item, to maintain it or to restore it to a state in which it can
accomplish its required function.
Maintenance, corrective
Corrective maintenance refers to actions performed on a failed item of equipment, to restore its
functional capability.
As the purpose of corrective maintenance is to correct functional failures, it is always fortuitous
in nature.
Maintenance, preventative
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Preventative maintenance refers to actions performed on an item of equipment with the aim of
reducing the likelihood of functional failure. It aims to prevent failures and therefore to achieve a
desired availability for the equipment over a fixed period. Preventative maintenance is always
scheduled (which does not mean that it is necessarily periodic) and can be either systematic or
conditional.
Maximum historically probable earthquake (SMHV)
This is defined as the most severe earthquake that is likely to have occurred over a period
comparable to the historical period, that is the previous 1,000 years.
Mechanical stability
Mechanical stability is the ability of a component to resist loads tending to modify its alignment
or positioning (for example by causing it to fall, topple,, slip or to shearing off parts). The
mechanical stability of a component requires the strength and stability of its supports
Molten Corium-Concrete Interaction [MCCI]
Concrete ablation / decomposition following direct contact with molten corium.
Monitored area
Area in which workers are likely to receive an annual dose between 1 and 6 mSv under normal
working conditions.
Normal operation
Normal operation refers to the state of a nuclear plant when it is within its normal operational
range. This includes normal power operation, shutting down starting-up, maintenance, testing
and refuelling.
Nuclide
A nuclide is a nuclear species characterized by a certain number of protons and neutrons that
every atomic nucleus of this species contains. Two atoms of the same nuclide therefore have
the same atomic number and same mass number.
Operability
Operability is the capability of an active component and its auxiliary and electrical supporting
systems to fulfil the functions essential for achieve its safety objective.
Operating procedures
The set of written and/or electronic documents specifying the tasks to be performed in order to
attain the required functional objectives during normal and abnormal unit operation
PCC-1 Normal operating events
These are events occurring during operation of the nuclear plant in its normal operational range,
including operation at normal power levels, shutdown operation, and operation during start-up,
maintenance, testing and reloading phases.
PCC-2 Anticipated Operational Occurrences
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These are anticipated operational events in which there is a departure from normal operating
(PCC-1) conditions, and which might be expected to occur one or more times during the lifetime
of the unit. Design provisions ensure that such processes should not cause significant damage
to the safety significant equipment or fission product barriers, or escalate into more serious
category PCC-3, PCC-4 or RRC events.
PCC-3
Reference incidents
These are initiating events that are more severe than PCC-2 transients, whose frequency of
occurrence is low enough that they would not be expected to occur during the lifetime of a
single unit, but for which the possibility of occurrence cannot be excluded during the lifetime of a
group of similar plants.
PCC-4
Reference accidents
These are initiating events considered in the plant design basis whose frequency of occurrence
is so low that they would not be expected to occur during the lifetime of any plant in a group of
similar units.
Peak stress
The peak stress at a given point is the difference between the total stress and the stress
corresponding to the linear distribution with the same moment and the same average value.
The basic characteristic of a peak stress is that it cannot cause any general distortion. Peak
Stress is therefore taken into account only if fatigue or fast fracture is considered. In fact, it is
the total stress at a given point resulting from all applied loads which are taken into account in
the determination of resistance to fatigue and fast fracture.
Plant condition: categories PCC-1 to PCC-4
The PCCs are initiating events which are used as the design basis for the safety systems,
structures and components that are provided to reduce the radiological risk due to the EPR
installation. The PCCs are grouped into Categories 1-4 depending on their anticipated
frequency of occurrence, as follows:
PCC-1:
Normal operation events
PCC-2:
Transients events
PCC-3:
Incidents events
PCC-4:
Accidents events
Plant power supply system
The set of items necessary for the supply of power to the plant’s electrical equipment between
the high voltage terminals of the step-down [unit] and auxiliary [station] transformers (HTA,
50kV) and the terminals of the receivers. The normal power supply system is divided into
normal and emergency electrical supply systems.
Power supply system, emergency
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The emergency electrical supply system is part of the plant’s electrical system and includes all
the electrical systems and equipment necessary to supply the power-consuming equipment in
carrying out the safety functions (e.g. establishment of a safe shutdown state, removal of
residual heat and prevention of radioactive release). Power for the emergency electrical supply
system is supplied by the emergency diesel generators in the event of unavailability of the
normal power supply system.
Power supply system, off-site
The off-site power supply system contains the main and auxiliary systems grid connections.
Power supply system, on-site
The on-site power supply system refers to all the power sources installed inside the plant
(except for the main alternator) that are available to supply emergency power to the plant
engineered safety features.
Power supply system, normal
The normal power supply system is part of the plant’s electrical system. It includes all the
electrical systems and equipment necessary for the supply to the auxiliaries.
Practical elimination
Practical elimination is a demonstration showing that an accident which could potentially lead to
large early releases of radioactive material to atmosphere can be ruled out from the list of
severe accidents addressed in the unit design. The practical elimination of accident situations
involves judgment and a separate examination of each type of sequence. Practical elimination
of accident sequences may involve deterministic and/or probabilistic considerations, and takes
into account uncertainties due to the limited knowledge of certain physical phenomena. It is
emphasized that practical elimination cannot be demonstrated simply by applying a generic
probabilistic cut-off value. (See Exclusion)
Primary stress
Primary stresses are the category of stresses that contribute directly to satisfying equilibrium of
mechanical loads. For this reason, they continue to exist in the event of plastic deformation and
a further increase of the external loads will cause a significant increase in deformation which is
not self-limiting. When the primary stresses exceed the yield strength of the material, there is a
risk of excessive deformation.
Protected evacuation route
A protected evacuation route is a transit area protected from the effects of fire. Such routes are
principally protected corridors or protected stairwells.
Protected rescue route
A protected rescue route is an area protected against the effects of a fire. Protected rescue
routes are protected corridors or stairwells.
Qualification
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A process of verification ensuring that an item of equipment required to meet certain
performance demands will function correctly and sufficiently reliably on demand, taking into
account the environmental conditions to which it will be exposed, including severe accident
conditions.
Quenching
The rapid cool down of a hot liquid or solid by immersion into or contact with water
Radioactive period (or half-life) of a radioactive nuclide
The half-life is the period at the end of which half of the initial quantity of a radionuclide has
disintegrated. It is a fixed characteristic of a radionuclide.
Radioactivity
The phenomenon of spontaneous transformation of a nuclide with the emission of ionizing
radiation.
Radionuclide
A radionuclide is a radioactive nuclide.
Reactor Coolant Pressure Boundary (CPP) [RCPB]
The Reactor Coolant Pressure Boundary consists of the primary circuit and all connecting lines
up to and including the second isolation valve or the second pilot-operated safety valve.
Reactor states, standard
In defining initiating events considered in the design basis for the unit, standard reactor states
A- F are defined as follows:
−
State A: at-power condition or hot or intermediate shutdown condition with all the
reactor’s automatic protection functions available. (Certain functions can be
unavailable at low pressure),
−
State B: intermediate shutdown above 120° C, shutdown cooling system not
connected. (Certain of the reactor’s automatic protection functions can be
unavailable),
−
State C: intermediate shutdown and cold shutdown with the shutdown cooling
system operational and an intact primary circuit or capable of being quickly made
intact,
−
State D: cold shutdown with primary circuit open (i.e. not intact),
−
State E: cold shutdown with filled reactor pool,
−
State F: cold shutdown with reactor core completely unloaded.
Redundancy
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The provision of a number of systems, structures or components (identical or diverse), so that
any one can perform the required function, regardless of the state of operation or failure of any
other.
Release target
Target for maximum fission product release into the environment allowable during a severe
accident. In general this target is initially set one order of magnitude below the value one strives
for to meet a certain environmental dose rate during a severe accident.
Risk Reduction: categories RRC-A and RRC-B
Event sequences involving multiple failures, which extend the design and safety assessment of
the NI against the Plant Initiating Events, are categorized into two Risk Reduction Categories.
These are:
RRC – A: Prevention of core melt
RRC – B: Prevention of large releases in core melt cases
(RRC-B sequences are referred to as severe accidents)
Sacrificial material
Sacrificial material refers to special concrete which is provided to be consumed after contact
with molten corium, in order to promote physical and chemical mechanisms aimed at ensuring
the long-term retention of the corium.
Safe shutdown earthquake (SMS) [SSE]
The intensity of the safe shutdown earthquake is taken as equal to that of the SMHV (maximum
historically probable earthquake) increased by one unit (MSK scale).
Safe shutdown State
For PCC 2-4 type events, the safe shutdown state is defined as:
−
−
−
core sub-critical,
decay heat removed,
activity release maintained at an acceptable level.
For PCC 2-4 type events affecting the spent-fuel pool, safe shutdown is defined as being a state
in which decay heat is being removed from the fuel stored in the pool by at least one train of
PTR [FPCS] cooling.
Scabbing
Scabbing is the ejection of irregularly shaped fragments due to degradation of a target area on
the side opposite to the point of impact of a missile.
Secondary stress
These are stresses that must be limited in order to ensure general structural adaptation.
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Secondary stresses are associated with the relative deformations in adjacent parts of a
component (or zone) analysed when the component (or zone) is subjected to mechanical loads
or thermal expansion. Secondary stresses are characterised by the production of self-limiting
plastic deformations when the yield stress is exceeded.
Segregation
The physical separation of components and systems, by distance or by some form of barrier,
that reduces the likelihood of common cause failure.
Separation, geographical
Achievement of geographical separation of two items of equipment requires installing at a
distance, or aligning them sufficiently differently, that their simultaneous loss in the event of
hazard or an initiating event need not be considered. The value of ‘sufficient separation
distance’ depends on the hazard or initiating event [PIE] under consideration.
Separation, physical
Physical separation of two items of equipment refers to their separation by means of an
appropriate barrier (e.g. a wall).
Severe accident
A severe accident or RCC-B sequence refers to any sequence leading to at least partial core
meltdown, which could consequently result in a large radiological release to the environment.
Single failure
A single failure is postulated as an independent failure and results in the loss of capability of a
component to perform its intended safety function.
Source term
The source term is the inventory of radioactive products inside and outside the containment,
which must be considered in the calculation of the radiological consequences of accidents.
Stratification
Stratification describes the phenomenon of local separation of gaseous or non-miscible liquid
components into regions/layers of different compositions and temperatures.
Sub-assembly
A sub-assembly is a part of a component composed of elementary parts.
Support
Supports are metal structures or non-integrated pieces of them. Movable supports allow relative
motion between the supported equipment and the supporting structure in the directions of the
action. Rigid supports lead the action from the equipment into the construction works.
They include for example: metal structures with support functions for mechanical equipment,
pipe whip restraints, hangers and dampers.
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Support, integral
Supports which are bolted, pinned or clamped to the pressure retaining component, or which
are mechanically attached to an integral mounting cast or forged integrally with the pressure
retaining component.
Supports which can directly support components.
Switchgear
The switchgear encompasses all the electrical connection components installed in cubicles,
including in particular: connection terminals, incoming and outgoing feeders, circuit breakers,
earthing switches, copper busbars, measuring devices, auxiliary relays, etc.
System
A system is a set of components which form a unit capable of performing specific functions
within the plant.
System, safety classified
A system is defined as being safety classified if it is classified according to functional
requirements (F1 or F2) or mechanical requirements (M1, M2 or M3).
Note: If only restricted parts of the system are classified according to functional or barrier
requirements (e.g. containment isolation, RCPB (CPP) isolation, secondary side isolation), the
system is not called safety classified.
System, protection
A protection system comprises all the electrical and mechanical devices and circuitry, from
sensors to actuation device, input terminals, involved in the monitoring of safety parameters and
generating signals associated with the protective function, whatever the state of the unit.
Systems or equipment, operational
The operational systems or equipment refer to the systems or equipment that are used for the
normal operation of the plant and are not safety classified (NC).
System, support
A support system enables the main function of a safety or operating system, for example:
electric power supply, cooling, lubrication or control.
System software
System software is software designed for a specific computer based system or for a family of
computer systems to facilitate operation and maintenance of the computer system and of its
associated programmes.
Thermal and mechanical loadings
Mechanical and thermal loadings are forces and moments, imposed deformations and nonuniform temperature distributions, in so far as they induce stresses and strains within a
component.
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Mechanical loadings (including thermal expansion loadings) can produce primary or secondary
stresses; thermal loadings only produce secondary stresses.
Total loss of internal and external electrical power supplies (Station Black Out)
This refers to a loss of the external (off-site) electrical power supplies combined with coincident
failure of all 4 emergency diesel generators, which results in a total loss of power to the
emergency and non-emergency 10kV busbars.
Transformer, auxiliary [station]
Transformer to supply power to the HTA (<50kV) electrical networks which enables plant
shutdown when the main grid and the alternator are unavailable.
Transformer, step down [unit]
Transformer to supply power to the HTA (<50kV) electrical networks in all operation states when
the main grid or the alternator are supplying a voltage within the admissible limits.
Transformer, main
Transformer used to adapt the main generator voltage to the high voltage (HTB) of the main
grid.
Type C (partial) leak resistance test
Type C (partial) leak resistance tests are tests designed to detect and measure localised leaks
by means of specific containment isolating valves.
Validation
The testing and evaluation of an integrated computer system (hardware and software) to ensure
compliance with its functional, performance and interface requirements.
Verification
A process which determines whether, at a given stage of development, a product fulfils all the
requirements set by the previous stage.
Voltage, nominal
The nominal voltage is the voltage between phases for which a network is designed.
Voltage, operational
The operational voltage is the adjusted voltage at the busbars during normal operation.
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3.
GLOSSARY OF ABBREVIATIONS
ABBREVIATIONS USED FOR SYSTEMS
Three-letter codes for the main systems
CODE FOR
ELEMENTARY
SYSTEM
AAD
EQUIVALENT
ENGLISH
ABBREVIATION
ABP
/
LOW PRESSURE FEEDWATER AND FEED HEATING
PLANT
ADG
/
FEEDWATER TANK
AHP
/
HIGH AND MEDIUM PRESSURE FEEDWATER AND
FEED HEATING PLANT
APA
MFWPS
APG
SGBS
STEAM GENERATOR BLOWDOWN SYSTEM
ARE
MFWS
STEAM GENERATORS MAIN FEEDWATER SYSTEM
ASG
EFWS
STEAM GENERATORS EMERGENCY FEEDWATER
SYSTEM
CET
/
TURBINE LABYRINTH SEAL
CEX
/
EXTRACTION CIRCUIT (RECOVERY PUMP)
CFI
CWFS
CRF
/
WATER CIRCULATION (LUBRICATION, FILTERING
AND SEPARATION)
CTE
/
CIRCULATING WATER TREATMENT
CVI
/
CONDENSER VACUUM
DCL
/
CONTROL ROOM AND ELECTRICAL EQUIPMENT
ROOMS ATMOSPHERIC CONDITIONING
DEL
/
ELECTRICAL BUILDING EMERGENCY CHILLED
WATER PRODUCTION
DEQ
/
EFFLUENT TREATMENT BUILDING CHILLED
WATER PRODUCTION
DER
/
REACTOR BUILDING CHILLED WATER
PRODUCTION
DFL
/
ELECTRICAL BUILDING FUME EXTRACTION
DMK
/
HANDLING EQUIPMENT AND PLANT FOR THE
FUEL BUILDING
DMR
/
HANDLING EQUIPMENT AND PLANT FOR THE
REACTOR BUILDING
DN.
/
NORMAL LIGHTING FOR BUILDINGS AND OPEN
AREAS SITE
SSS
DESCRIPTION
STARTUP AND SHUTDOWN FEED PUMP TRAIN
MAIN FEED PUMP TRAIN (INCLUDING
LUBRICATION)
SEA WATER FILTERING
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CODE FOR
ELEMENTARY
SYSTEM
DNX
EQUIVALENT
ENGLISH
ABBREVIATION
/
POWER OUTLET DISTRIBUTION
DS.
/
EMERGENCY LIGHTING FOR BUILDINGS AND
OPEN AREAS OF SITE
DVB
/
VENTILATION AND AIR-CONDITIONING OF
ADMINISTRATIVE OFFICES
DVD
/
VENTILATION AND HEATING OF DIESEL
BUILDINGS
DVL
/
VENTILATION OF ELECTRICAL ROOMS IN THE
ELECTRICAL BUILDING
DVP
/
VENTILATION OF THE PUMPING STATION
DWB
/
VENTILATION OF THE CONTAMINABLE ROOMS IN
THE OPERATIONAL PRODUCTION CENTRE
DWK
/
VENTILATION OF THE FUEL BUILDING
DWL
CSBVS
DWN
/
VENTILATION OF THE NUCLEAR AUXILIARY
BUILDING
DWQ
/
VENTILATION OF THE EFFLUENT TREATMENT
BUILDING
DWW
/
VENTILATION OF CONTAMINABLE ROOMS IN THE
ACCESS TOWER
EBA
CSVS
EDE
AVS
EPP
/
SEALING AND CONTROL OF CONTAINMENT LEAKS
(AIR LOCKS, BUSHINGS, BUFFERS, ETC.)
ETY
CGCS
CONTAINMENT H2 CONTROL (COMBUSTIBLE GAS
COMBUSTION CONTROL)
EVF
/
EVR
CCVS
CONTAINMENT COOLING AND VENTILATION
EVU
CHRS
CONTAINMENT HEAT REMOVAL
GCT
MSB
GEA
/
AUXILIARY TRANSFORMER
GEV
/
POWER TRANSMISSION SYSTEM (INCLUDING
STEP DOWN TRANFORMER)
GEX
/
ALTERNATOR EXCITATION AND REGULATION
GPV
/
MAIN TURBINE STEAM AND BLOWDOWN SYSTEM
GRE
/
TURBINE GOVERNING SYSTEM
GRV
/
ALTERNATOR H2 SUPPLY
GSE
/
TURBINE SAFETY DEVICES (PROTECTION)
GSS
/
MOISTURE SEPARATORS/REHEATERS
JAC
/
CLASSIFIED FIRE-FIGHTING WATER PRODUCTION
JDT
FDS
DESCRIPTION
VENTILATION FOR THE SAFEGUARD AUXILIARY
BUILDING AND THE ELECTRICAL BUILDING
PURGING OF THE REACTOR BUILDING
ANNULUS VENTILATION SYSTEM
INTERNAL CONTAINMENT FILTRATION
MAIN STEAM BYPASS (TO CONDENSER)
FIRE DETECTION
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CODE FOR
ELEMENTARY
SYSTEM
JPD
EQUIVALENT
ENGLISH
ABBREVIATION
/
CONVENTIONAL ISLAND FIRE-FIGHTING WATER
PROTECTION AND DISTRIBUTION
JPH
/
TURBINE HALL OIL TANKS FIRE-FIGHTING WATER
PROTECTION AND DISTRIBUTION
JPI
/
NUCLEAR ISLAND FIRE-FIGHTING WATER
PROTECTION AND DISTRIBUTION
JPS
/
SITE FIRE-FIGHTING WATER PROTECTION AND
DISTRIBUTION
JPT
/
TRANSFORMERS FIRE-FIGHTING WATER
PROTECTION AND DISTRIBUTION
JPV
/
DIESEL GENERATOR SETS FIRE-FIGHTING WATER
PROTECTION AND DISTRIBUTION
KER
LRMDS
CONTROL AND DISCHARGE OF EFFLUENTS FROM
THE NUCLEAR ISLAND
KIR
/
PRIMARY SYSTEM MONITORING
INSTRUMENTATION
KKK
/
GENERAL ACCESS CONTROL
KRC
/
BODILY CONTAMINATION AND DOSIMETRY
CONTROL AND IRRADIATION OF PREMISES
KRT
RPMS
LA.
/
PRODUCTION AND DISTRIBUTION OF
ELECTRICITY AT 220V DC
LAK/L/M/N
/
DC 220 V POWER SUPPLY CLUSTER CONTROL
MECHANISM
LAV/W
/
DC 220 V POWER SUPPLY GTA PLANT
AUXILIARIES AND CLUSTER CONTROL
MECHANISM
LG.
/
DISTRIBUTION > OR = TO 5.5 kV AC (NOT BACKED
UP)
LGA/B/C/D
/
NORMAL 10 kV DISTRIBUTION CONVENTIONAL
ISLAND
LGE/LGJ
/
NORMAL 10 kV DISTRIBUTION OPERATIONAL
PRODUCTION CENTRE (HB)
LGF/G/H/I
/
NORMAL 10 kV DISTRIBUTION NUCLEAR ISLAND
LGK/LGN
/
NORMAL 10 kV DISTRIBUTION CONVENTIONAL
ISLAND
LGP/Q/R/S
/
NORMAL 10 kV DISTRIBUTION CONVENTIONAL
ISLAND (PUMPING STATION)
LH.
/
PRODUCTION AND DISTRIBUTION AC > OR = 5.5 kV
BACKED UP
LHA/B/C/D
EPSS
BACKED UP 10 kV DISTRIBUTION (NUCLEAR
ISLAND + BACKUP) – EPSS: EMERGENCY POWER
SUPPLY SYSTEM
LHP/Q/R/S
EDG
10 kV DIESEL UNIT DIVISIONS 1/2/3/4 (EDG :
EMERGENCY DIESEL GENERATOR)
DESCRIPTION
RADIOLOGICAL PROTECTION MEASUREMENT
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CODE FOR
ELEMENTARY
SYSTEM
LI.
EQUIVALENT
ENGLISH
ABBREVIATION
/
690 V ALTERNATE NORMAL DISTRIBUTION
LIA/B/C/D
/
NORMAL 690 V DISTRIBUTION CONVENTIONAL
ISLAND
LIF/LII
/
NORMAL 690 V DISTRIBUTION NUCLEAR ISLAND
LJ.
/
PRODUCTION AND DISTRIBUTION OF 690 V AC
BACKED UP
LJA/B/C/D
/
690 V AC BACKED UP NUCLEAR ISLAND
LJF/LJI
/
690 V AC BACKED UP NUCLEAR ISLAND
LJL/M
/
690 V AC BACKED UP CONVENTIONAL ISLAND
LJP/S
SBO DIESELS
690 V EMERGENCY DIESEL GENERATOR SET
DIVISIONS 1 AND 4
LJU/X
/
690 V ALTERNATE BACKED UP NUCLEAR ISLAND,
DIESEL GENERATOR BUILDINGS
LJZ
/
690 V AC BACKED UP NUCLEAR ISLAND, THIRD
PTR PUMP
LK.
/
400 V AC NORMAL DISTRIBUTION (UNDER LG
BOARD)
LKA/B/C/D
/
NORMAL 400 V DISTRIBUTION CONVENTIONAL
ISLAND
LKF/G/H/I
/
NORMAL 400 V DISTRIBUTION CONVENTIONAL
ISLAND, PUMPING STATION
LKK/L/M/N
/
NORMAL 400 V DISTRIBUTION NUCLEAR ISLAND
LKP/Q/R/S
/
NORMAL 400 V DISTRIBUTION NUCLEAR ISLAND
LL.
/
400 V AC BACKED UP DISTRIBUTION (UNDER LH
BOARD)
LLA/B/C/D
/
BACKED UP 400 V DISTRIBUTION NUCLEAR
ISLAND
LLF/G/H/I
/
BACKED UP 400 V DISTRIBUTION NUCLEAR
ISLAND (LH DIESEL AUXILIARIES)
LLL/LLM
/
BACKED UP 400 V DISTRIBUTION CONVENTIONAL
ISLAND
LLP/Q/R/S
/
BACKED UP 400 V DISTRIBUTION NUCLEAR
ISLAND
LO.
/
400 V AC REGULATED DISTRIBUTION
LOA/B/C/D
/
400 V REGULATED DISTRIBUTION NUCLEAR
ISLAND
LOF/G/H/I
/
400 V REGULATED DISTRIBUTION NUCLEAR
ISLAND (CONTAINMENT ISOLATION)
LTR
/
EARTH CIRCUIT
LV.
/
UNINTERRUPTED 400 V PRODUCTION AND
DISTRIBUTION
LVA/B/C/D
/
UNINTERRUPTED 400 V PRODUCTION AND
DISTRIBUTION NUCLEAR ISLAND
DESCRIPTION
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CODE FOR
ELEMENTARY
SYSTEM
LVF/G/H/I
EQUIVALENT
ENGLISH
ABBREVIATION
/
UNINTERRUPTED 400 V PRODUCTION AND
DISTRIBUTION NUCLEAR ISLAND
LVL/M
/
UNINTERRUPTED 400 V PRODUCTION AND
DISTRIBUTION CONVENTIONAL ISLAND
LVP/S
/
UNINTERRUPTED 400 V PRODUCTION AND
DISTRIBUTION NUCLEAR ISLAND
PMA
/
FUEL HANDLING AUXILIARIES
PMC
/
FUEL HANDLING – LOADING MACHINE –
OVERHEAD CRANE – TRANSFER
PMG
/
MULTISTUD TENSIONING MACHINE (MSDG)
PMO
/
HANDLING TOOLS - LIGHTING – ASPIRATOR
REACTOR BUILDING PLATFORM
PTR
FPPS/
FPCS
DESCRIPTION
FUEL POOL PURIFICATION AND COOLING
SYSTEMS
RBS
EBS
EXTRA BORATION SYSTEM
RCP
RCS
REACTOR PRIMARY COOLANT
RCV
CVCS
REA
RBWMS
REN
NSS
RES
/
RGL
CRDM
RIC
/
RIS
SIS
RPE
NVDS
RPN
/
RPR
PS OR RPS
RRI
CCWS
SA
/
PRODUCTION AND DISTRIBUTION OF AIR
SAA
/
PRODUCTION OF BREATHABLE AIR
SAP
/
PRODUCTION OF COMPRESSED
INSTRUMENT/SERVICE AIR
SAR
/
INSTRUMENT COMPRESSED AIR DISTRIBUTION
SAT
/
SERVICE COMPRESSED AIR DISTRIBUTION
SDA
/
DEMINERALIZED WATER PRODUCTION
SDS
/
PRODUCTION OF DEMINERALIZED WATER BY
DESALINATION
SEA
/
WATER DEMINERALIZATION (PRE-TREATMENT)
SEC
ESWS
CHEMICAL AND VOLUME CONTROL
REACTOR BORON AND WATER MAKEUP
NUCLEAR SAMPLING
STEAM GENERATOR SECONDARY SYSTEM
SAMPLING
CONTROL ROD DRIVE MECHANISM
CORE INTERNAL INSTRUMENTATION
SAFETY INJECTION SYSTEM
NUCLEAR VENTS AND DRAINS
NUCLEAR POWER MEASUREMENT
REACTOR PROTECTION SYSTEM
COMPONENT COOLING WATER
ESSENTIAL SERVICES WATER SYSTEM (RRI
COMPONENT COOLING SYSTEM)
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CODE FOR
ELEMENTARY
SYSTEM
SED
EQUIVALENT
ENGLISH
ABBREVIATION
/
DISTRIBUTION OF DEMINERALIZED REACTOR
WATER
SEF
/
WATER INTAKE – FILTERING – TRASH RAKES
SHE
/
COLLECTION OF OILS AND HYDROCARBON
EFFLUENTS (INCLUDING STORAGE)
SEK
CILWDS
LWDS site
SEL
/
DISTRIBUTION OF ELECTRICALLY-HEATED HOT
WATER
SEN
/
SRI RAW WATER COOLING
SEO
/
WATER LOST IN SEWER
SEP
/
DRINKING WATER
SER
/
PH9 CONVENTIONAL DEMINERALIZED WATER
DISTRIBUTION INSTALLATION (STORAGE
INCLUDED)
SG.
/
DISTRIBUTION OF FLUIDS OTHER THAN WATER
AND AIR
SGC
/
DISTRIBUTION OF CARBON DIOXIDE
SGH
/
DISTRIBUTION OF HYDROGEN
SGN
/
DISTRIBUTION OF NITROGEN
SGO
/
DISTRIBUTION OF OXYGEN
SIR
/
CHEMICAL CONDITIONING (INJECTION WITH
REAGENTS)
SKZ
/
STORAGE OF GASES (H2, O2, N2, CO2 AND RARE
GASES)
SNL
/
STEAM GENERATOR UNIT CLEANING – LANCING
SRI
/
COMPONENT COOLING – CONVENTIONAL
SYSTEMS (NORIA)
SRU
UCWS
SVA
/
AUXILIARY STEAM DISTRIBUTION
TEG
GWPS
GASEOUS WASTE PROCESSING
TEN
/
TEP
CSTS
PRIMARY LIQUID EFFLUENT
TER
ExLWDS
RESIDUAL LIQUID EFFLUENT
TES
SWTS
SOLID WASTES TREATMENT
TEU
LWPS
NON-RECYCLED LIQUID WASTE TREATMENT
TRI
/
VDA
MSRT
MAIN STEAM ATMOSPHERIC DUMP (MAIN STEAM
RELIEF TRAINS)
VIV
MSIV
MAIN STEAM ISOLATION VALVE
DESCRIPTION
COLLECTION, CONTROL AND DISCHARGE OF
SECONDARY SYSTEM EFFLUENTS
ULTIMATE COOLING
SYSTEMS FOR SAMPLING EFFLUENT FROM
EFFLUENT TREATMENT BUILDING
COMPONENT COOLING EFFLUENT TREATMENT
CHAPTER: A
FUNDAMENTAL SAFETY OVERVIEW
UK-EPR
VOLUME 2: DESIGN AND SAFETY REPORT
CHAPTER A: INTRODUCTION AND GLOSSARY
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CODE FOR
ELEMENTARY
SYSTEM
VPU
EQUIVALENT
ENGLISH
ABBREVIATION
VVP
MSSS
/
DESCRIPTION
STEAM LINE DRAIN SYSTEM
MAIN STEAM SYSTEM, VENTS VALVES AND
STEAM GENERATOR VENTS AND VALVES
Other abbreviations used for systems
FRENCH
ABBREVIATION
ENGLISH
ABBREVIATION
DEA
SSSS
STANDSTILL SEAL SYSTEM
IRWST
or RIS pool
IRWST
IN-CONTAINMENT REFUELING WATER STORAGE
TANK
ISBP
LHSI
LOW-HEAD SAFETY INJECTION
ISMP
MHSI
MEDIUM-HEAD SAFETY INJECTION
MCP
PICS
PROCESS INFORMATION AND CONTROL
MCS
SICS OR PICS
SAFETY INFORMATION AND CONTROL
MSRIV
MAIN STEAM RELIEF ISOLATION VALVE (PART OF
VDA SYSTEM, I.E. ATMOSPHERIC DUMP SYSTEM)
MSRV
MAIN STEAM RELIEF VALVE (PART OF THE VDA
SYSTEM, I.E. ATMOSPHERIC DUMP SYSTEM)
MSS
MAIN STEAM SYSTEM
MSSV
MAIN STEAM SAFETY RELIEF VALVE
NPSS
NORMAL POWER SUPPLY SYSTEM
PROCESS CONTROL SYSTEM
PAS
PACS
DESCRIPTION
PACS
PRIORITY AND ACTUATOR CONTROL SYSTEM
PAMS
POST-ACCIDENT MONITORING SYSTEM
PAS
PROCESS AUTOMATION SYSTEM
PDS
PRESSURISER DEPRESSURISATION SYSTEM
PS
RPS
REACTOR PROTECTION SYSTEM
RCSL
RCSL
RDP
PRT
PRESSURISER RELIEF TANK
RIS-RA
RHR
RESIDUAL HEAT REMOVAL
SAS
SAS
SAFETY AUTOMATION SYSTEM
SDC
MCR
MAIN CONTROL ROOM
SDR
RSS
REMOTE SHUTDOWN STATION
SYN
POP
MIMIC BUS OR PLANT OVERVIEW PANEL
TXS
TXS
TELEPERM XS
PAS
REACTOR CONTROL, SURVEILLANCE AND
LIMITATION SYSTEM
CHAPTER: A
FUNDAMENTAL SAFETY OVERVIEW
UK-EPR
VOLUME 2: DESIGN AND SAFETY REPORT
CHAPTER A: INTRODUCTION AND GLOSSARY
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ABBREVIATIONS FOR MAIN BUILDINGS AND TUNNELS
COMMON FRENCH
ABBREVIATION
EDF CODING
BAN
HN
BAS
HLF, HLG
HLH, HLI
SAFEGUARD BUILDING
BD
HDA, HDB
HDC, HDD
DIESEL BUILDINGS
BK
HK
BL
HLA, HLB
HLC, HLD
BLNC
HF
UNCLASSIFIED ELECTRICAL BUILDING
BR
HR
REACTOR BUILDING
BTE
HQA, HQB
BZ
HZ
GAS AND CHEMICAL PRODUCTS STORAGE
BUILDING
POE
HB
OPERATIONAL SERVICE CENTRE
SDM
HM
TURBINE HALL
SDP
HP
TA
HJ
PUMPING STATION (INCLUDING DISCHARGE
STRUCTURES)
AUXILIARY TRANSFORMER PLATFORM
TP-TS
HT
/
HCA, HCB
/
HW
ACCESS TOWER BUILDING
/
HX
EFFLUENT STORAGE AREA
HY
DEMINERALIZATION BUILDING
DESCRIPTION
NUCLEAR AUXILIARIES BUILDING
FUEL BUILDING
ELECTRICAL BUILDING
EFFLUENT TREATMENT BUILDING
MAIN AND STEP DOWN TRANSFORMER
PLATFORM
DISCHARGE AND PRE-DISCHARGE PONDS
DESALINATION UNIT
/
HYS
COMMON FRENCH
ABBREVIATION
EDF CODING
/
HGB
ACCESS TUNNELS TO OPERATIONAL SERVICE
CENTRE
/
HGD
DIESEL – BAS TUNNELS
/
HGM/HGL
/
HGN
BAN MECHANICAL/ELECTRICAL TUNNEL
/
HGP
TURBINE HALL -PUMPING STATION TUNNEL
/
HGQ
BAN – BTE TUNNEL
/
HGS
BAS - PUMPING STATION TUNNEL (EMERGENCY)
/
HGT
UNCLASSIFIED ELECTRICAL BUILDING –
TRANSFORMER TUNNEL
DESCRIPTION
MECHANICAL/ELECTRICAL TUNNEL
CHAPTER: A
FUNDAMENTAL SAFETY OVERVIEW
UK-EPR
VOLUME 2: DESIGN AND SAFETY REPORT
CHAPTER A: INTRODUCTION AND GLOSSARY
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COMMON FRENCH
ABBREVIATION
EDF CODING
/
HGW
DESCRIPTION
OPERATIONAL SERVICE CENTRE – ACCESS
TOWER TUNNEL
OTHER ABBREVIATIONS (ENGLISH ABBREVIATION NOT GIVEN
IF NOT USED OR SAME AS THE SO-CALLED COMMON
ABBREVIATION)
COMMON
ABBREVIATION
ENGLISH
ABBREVIATION
DESCRIPTION
AAC
HOT SHUTDOWN
AAF
COLD SHUTDOWN
AAR
AUTOMATIC REACTOR SHUTDOWN
AC
ALTERNATING CURRENT
AG
SEVERE ACCIDENT
-
AP
AIEA
IAEA
ACTIVATION PRODUCT
INTERNATIONAL ATOMIC ENERGY AGENCY
ALARA
AS LOW AS REASONABLY ACHIEVABLE
ALARP
AS LOW AS REASONABLY PRACTICABLE
APE
STATE-ORIENTED APPROACH
API
SHUTDOWN FOR MAINTENANCE
APR
SHUTDOWN FOR FUEL RELOADING
APRP
LOCA
AQ
QA
LOSS OF PRIMARY COOLANT ACCIDENT
QUALITY ASSURANCE
AN/GV
NORMAL SHUTDOWN WITH HEAT REMOVAL BY
STEAM GENERATORS
AN/RRA
NORMAL SHUTDOWN WITH HEAT REMOVAL BY
RIS/RRA
AO
AXIAL OFFSET
ASR
SHUTDOWN FOR SIMPLE RELOADING
ATWS
ANTICIPATED TRANSIENT WITHOUT SCRAM
ATWT
ANTICIPATED TRANSIENT WITHOUT TRIP
BAE
BC
WATER BOX
HL
BDOP
BASIC DESIGN OPTIMIZATION PHASE
BDR 99
BF
BI
HOT LEG
BASIC DESIGN REPORT (EDITION 1999)
CL
COLD LEG
INTERMEDIATE LEG
CHAPTER: A
FUNDAMENTAL SAFETY OVERVIEW
UK-EPR
VOLUME 2: DESIGN AND SAFETY REPORT
CHAPTER A: INTRODUCTION AND GLOSSARY
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COMMON
ABBREVIATION
ENGLISH
ABBREVIATION
DESCRIPTION
BNI
NUCLEAR ISLAND APART FROM NUCLEAR STEAM
SUPPLY SYSTEM (BALANCE OF NUCLEAR
ISLAND)
BOP
BALANCE OF PLANT (PLANT OUTSIDE
CONVENTIONAL AND NUCLEAR ISLANDS)
CAD
CAD
CB
CC
COMPUTER-AIDED DESIGN
BORON CONCENTRATION
I&C
CDF
INSTRUMENTATION AND CONTROL SYSTEM
CORE DAMAGE FREQUENCY
CDU
SFC
SINGLE FAILURE CRITERION
CIPR
ICRP
INTERNATIONAL COMMISSION ON RADIOLOGICAL
PROTECTION
CNEN
NATIONAL CENTRE FOR NUCLEAR EQUIPMENT
CNEPE
NATIONAL CENTRE FOR NUCLEAR PRODUCTION
EQUIPMENT
CNPE
CPP
NATIONAL CENTRE FOR ELECTRICITY
PRODUCTION
RCPB
CREDOC
CSP
DAC
RESEARCH CENTRE FOR THE STUDY AND
OBSERVATION OF LIVING CONDITIONS
SSPB
SECONDARY SYSTEM PRESSURE BOUNDARY
CREATION AUTHORISATION REQUEST
DARPE
AUTHORISATION REQUEST FOR WATER
EXTRACTION AND DISCHARGES
DBC
DESIGN BASED CONDITION
DC
DCC
REACTOR COOLANT PRESSURE BOUNDARY
DIRECT CURRENT
CCF
DCH
COMMON CAUSE FAILURE
DIRECT CONTAINMENT HEATING
DDC
BOC
BEGINNING OF CYCLE
DDV
BOL
BEGINNING OF LIFE
DGSNR
DN
DNB(R)
FRENCH GENERAL DIRECTORATE FOR NUCLEAR
SAFETY AND RADIOLOGICAL PROTECTION
(DIRECTION GENERALE DE LA SURETE
NUCLEAIRE ET DE LA RADIOPROTECTION)
NOMINAL DIAMETER
DEPARTURE FROM NUCLEATE BOILING (RATIO)
DT
TURBINE TRIP
DT
TECHNICAL DIRECTIVES
EDF
ELECTRICITE DE FRANCE
ECS
EDF CODING SYSTEM
EDS
ELECTRICAL DISTRIBUTION SYSTEM
CHAPTER: A
FUNDAMENTAL SAFETY OVERVIEW
UK-EPR
VOLUME 2: DESIGN AND SAFETY REPORT
CHAPTER A: INTRODUCTION AND GLOSSARY
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COMMON
ABBREVIATION
ENGLISH
ABBREVIATION
EP
DESCRIPTION
PERIODIC TESTS
EPR
EUROPEAN PRESSURIZED WATER REACTOR
EPRI
ELECTRIC POWER RESEARCH INSTITUTE (USA)
EPS
PSA
PROBABILISTIC SAFETY ANALYSES
EPW
EXPLOSION PRESSURE WAVE
ESP
PRESSURE-RETAINING BOUNDARY
ESPN
TECHNICAL CODE FOR PRESSURE RETAINING
BOUNDARY FOR NUCLEAR EQUIPMENT
ETC-C
EPR TECHNICAL CODE FOR CIVIL WORKS
ETC-F
EPR TECHNICAL CODE FOR FIRE PROTECTION
EUR
EUROPEAN UTILITY REQUIREMENTS
FAC
CHAIN FILTER
FAR
LBB
LEAK BEFORE RUPTURE
FDC
EOC
END OF CYCLE
FDV
EOL
END OF LIFE
FH
HF
HUMAN FACTOR
-
FP
FISSION PRODUCT
GEMMES
-
GMPP
RCP
GPR
GV
REACTOR COOLANT PUMP
REACTOR SAFETY ADVISORY GROUP OF DGSNR
SG
HEPA
HTA
FUEL MANAGEMENT SCHEME FOR 1300MWE
ENRICHED UO2 UNITS
STEAM GENERATOR
HIGH EFFICIENCY PARTICULATE AIR (FILTER)
HVA
HTB
HIGH VOLTAGE (UP TO 50 kV)
HIGH VOLTAGE (OVER 50 kV)
HVAC
HEATING VENTILATION AND AIR CONDITIONING
I&C
C&I
INSTRUMENTATION AND CONTROL
ICB
CCI
CORIUM-CONCRETE INTERACTION
ICPE
CLASSIFIED INSTALLATION FOR ENVIRONMENTAL
PROTECTION
IEG
GENERAL ELECTRICAL LAYOUT
IFH
HFE
IJPP
RCPSWI
IHM
HMI OR MMI
HUMAN FACTORS ENGINEERING
RCP SEAL INJECTION
HUMAN-MACHINE INTERFACE
INB
BASIC NUCLEAR INSTALLATION (INSTALLATION
NUCLEAIRE DE BASE)
INSAG
INTERNATIONAL NUCLEAR SAFETY ADVISORY
GROUP
CHAPTER: A
FUNDAMENTAL SAFETY OVERVIEW
UK-EPR
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CHAPTER A: INTRODUCTION AND GLOSSARY
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COMMON
ABBREVIATION
ENGLISH
ABBREVIATION
IPG
PCI
IRSN
JEPP
DESCRIPTION
PELLET-CLADDING INTERACTION
FRENCH INSTITUTE FOR RADIOLOGICAL
PROTECTION AND NUCLEAR SAFETY (INSTITUT
DE LA RADIOPROTECTION ET DE SURETE
NUCLEAIRE)
EFPD
EQUIVALENT FULL POWER DAYS
KD
AVAILABILITY FACTOR
LBB
LEAK BEFORE BREAK
LBLOCA
LARGE BREAK LOSS OF COOLANT ACCIDENT
LCO
LIMITING CONDITION OF OPERATION
LERF
LARGE EARLY RELEASE FREQUENCY
LOCC
LOSS OF SUPPLY CHAIN COOLING
LOFW
LOSS OF SG FEEDWATER
LRF
LARGE RELEASE FREQUENCY
LSSS
LTC
LIMTING SAFETY SYSTEM SETTING
TSC
LUHS
TECHNICAL SUPPORT CENTRE
LOSS OF ULTIMATE HEAT SINK
MDNBR
MINIMUM DEPARTURE FROM NUCLEATE BOILING
RATIO
MDTE
LOOP
MDTG
SBO
LOSS OF OFF-SITE VOLTAGE
GENERALISED LOSS OF VOLTAGE OR STATION
BLACK-OUT
MEL
MASS AND ENERGY RELEASED
MOX
MIXED OXIDE FUEL
MQ
QM
QUALITY MANUAL
MSDG
MST
MULTISTUD TENSIONER
MSI
INDUSTRIAL COMMISSIONING
NC
NON SAFETY CLASSIFIED
NPSH
OEF
NET POSITIVE SUCTION HEAD
OEF
OPERATING EXPERIENCE FEEDBACK
OPP
OVERPRESSURE PROTECTION
PCC
PLANT CONDITION CATEGORIES (See PCC
Definition)
PDD
POWER DENSITY DETECTOR
PJC
TANK SEALING SURFACES
PN
NOMINAL POWER
PPI
SPECIAL EMERGENCY PLAN
PT
PARTIAL TRIP
PTAEE
LOOP
LOSS OF OFF-SITE POWER
CHAPTER: A
FUNDAMENTAL SAFETY OVERVIEW
UK-EPR
VOLUME 2: DESIGN AND SAFETY REPORT
CHAPTER A: INTRODUCTION AND GLOSSARY
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COMMON
ABBREVIATION
ENGLISH
ABBREVIATION
PTEA
TLOFW
PUI
PZR
TOTAL LOSS OF SG FEEDWATER
ON-SITE EMERGENCY PLAN
PZR
R&D
RAP
DESCRIPTION
PRESSURISER
RESEARCH & DEVELOPMENT
PAR
RAZ
PASSIVE AUTOCATALYTIC RECOMBINER
RESET
RCCA
ROD CLUSTER CONTROL ASSEMBLY
RCC-E
DESIGN AND CONSTRUCTION RULES
APPLICABLE TO ELECTRICAL EQUIPMENT AT A
NUCLEAR SITE
RCC-M
DESIGN AND CONSTRUCTION RULES
APPLICABLE TO MECHANICAL EQUIPMENT AT A
NUCLEAR SITE
RCD
CORE COMPLETELY UNLOADED
RDP
PRT
PRESSURISER RELIEF TANK
REP
PWR
PRESSURISED-WATER REACTOR
RFS
RFTC
BASIC SAFETY RULE
CHFR
RIGZ
RPS
CRITICAL HEAT FLUX RATIO
UNCONTROLLED WITHDRAWAL OF CONTROL
ROD GROUP AT ZERO POWER
PSR
PRELIMINARY SAFETY REPORT
RPV
REACTOR PRESSURE VESSEL
RRC
RISK REDUCTION CATEGORIES (See RRC
Definition)
RTE
FWLB
FEEDWATER LINE BREAK
RTGV
SGTR
STEAM GENERATOR TUBE RUPTURE
RTHE
HELB
HIGH ENERGY PIPING BREAKAGE
RTV
MSLB
MAIN STEAM LINE BREAK
SBLOCA
SDD
SMALL BREAK LOSS OF COOLANT ACCIDENT
DBE
SE
SMS
ELEMENTARY SYSTEM
SSE
SPND
STE
DESIGN BASIS EARTHQUAKE
SAFE SHUTDOWN EARTHQUAKE
SELF-POWERED NEUTRON DETECTOR
TS
TDD
OPERATING TECHNICAL SPECIFICATIONS
DEFLAGRATION DETONATION TRANSITION
TGE
ARO
ALL RODS OUT
TGI
ARI
ALL RODS IN
TLOCC
TRIC
TOTAL LOSS OF COMPONENT COOLING
COTC
CORE OUTLET TEMPERATURE
CHAPTER: A
FUNDAMENTAL SAFETY OVERVIEW
UK-EPR
VOLUME 2: DESIGN AND SAFETY REPORT
CHAPTER A: INTRODUCTION AND GLOSSARY
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COMMON
ABBREVIATION
ENGLISH
ABBREVIATION
DESCRIPTION
UCP
UPPER CORE PLATE
UPS
UNINTERRUPTIBLE POWER SUPPLY
USP
UPPER SUPPORT PLATE
VCT
VOLUME CONTROL TANK
VD
10-YEARLY IN-SERVICE INSPECTION
VP
OUTAGE INSPECTION