Glossary


A

Abnormal Operational Procedure
AOP Event based procedures applicable for more complex AOOs, typically events impacting the operation of more than one system or component.
Accident Management
AM or A/M The taking of a set of actions during the evolution of an accident:
- To prevent escalation to a severe accident;
- To mitigate the consequences of a severe accident;
- To achieve a long term safe stable state.
 
Alarm Response Procedures
ARP Procedures applied, whether available, in case of simple cases of AOOs (malfunctions).
 
Alternating Current
AC In alternating current, the flow of electric charge periodically reverses direction.
 
American Society of Mechanical Engineers
ASME American Society of Mechanical Engineers.
ASME III: Rules for Construction of Nuclear Facility Components.
 
Anticipated Operational Occurrence
AOO A deviation of an operational process from normal operation that is expected to occur at least once during the operating lifetime of a facility but which, in view of appropriate design provisions, does not cause any significant damage to items important to safety or lead to accident conditions.
 
Anticipated Transient Without Scram
ATWS For a nuclear reactor, an accident for which the initiating event is an anticipated operational occurrence and in which the system for fast shutdown of the reactor fails to function.

 
As Low As Reasonably Achievable
ALARA As low as reasonably achievable, economic and social factors being taken into account. ALARA means making every reasonable effort to maintain exposures to ionizing radiation as far below the dose limits as practical, consistent with the purpose for which the licensed activity is undertaken, taking into account the state of technology, the economics of improvements in relation to state of technology, the economics of improvements in relation to benefits to the public health and safety, and other societal and socioeconomic considerations, and in relation to utilization of nuclear energy and licensed materials in the public interest.

 
Auxiliary Feedwater (AFW) System
AFW The AFW is used in all pressurized water reactor designs. The main purpose of the AFW system is to provide feedwater to the steam generators to maintain a heat sink in the event of (i) a loss of main feedwater, (ii) a reactor trip and loss of offsite power, (iii) a small break loss of coolant accident. The system, at some plants, can also provide a source of feedwater to the steam generators during plant startup and shutdown.
 
Average Power Range Monitor
APRM The Average Power Range Monitoring System receives inputs from the Local Power Range Monitoring (LPRM) System to determine the average core power.

 

B

Barrier
A physical obstruction that prevents or inhibits the movement of people, radionuclides or some other phenomenon (e.g. fire), or provides shielding against radiation.

 
Beyond Design Basis Accident
BDBA Postulated accident with accident conditions more severe than those of a design basis accident.

 
Boiling Water Reactor
BWR Boiling light water cooled and moderated reactor. In a BWR, the reactor core heats water, which turns to steam and then drives a steam turbine.

 

C

Candidate High Level Actions
CHLA General actions that could be taken to manage a severe accident. The CHLAs formed the basis of generic guidance developed by the various owners groups representing the nuclear steam supply system (NSSS) vendors. This generic guidance was ultimately used to assemble the plant-specific guidance for each operating nuclear power plant.

 
CANDU reactor
CANDU CANDU is a trade mark of Atomic Energy of Canada Ltd., a nuclear reactor in which the fuel is natural uranium oxide clad in zircaloy and the coolant and moderator is heavy water.
 
Cladding
Typically, the tube of material that houses nuclear fuel pellets and provides the containment (means of confinement) of radionuclides produced during fission.

 
Computational Fluid Dynamics
CFD

 
Containment system
A structurally closed physical barrier (especially in a nuclear installation) designed to prevent or control the release and the dispersion of radioactive substances, and its associated systems.

 
Control rod
A rod, plate, or tube containing a material (such as hafnium, boron, etc.) used to control the power of a nuclear reactor. By absorbing neutrons, a control rod prevents the neutrons from causing further fissions.

 
Controlled state
Plant state, following an anticipated operational occurrence or accident conditions, in which fulfilment of the fundamental safety functions can be ensured and which can be maintained for a time sufficient to implement provisions to reach a safe state.

 
Coolant
A substance that is used to remove or transfer heat from the reactor core. Common coolants include light water, heavy water, air, carbon dioxide, helium, liquid sodium, and a sodium-potassium alloy.

 
Core Exit Temperature
CET
 

D

Decommissioning
Administrative and technical actions taken to allow the removal of some or all of the regulatory controls from a facility.

 
Design Basis
The range of conditions and events taken explicitly into account in the design of structures, systems and components and equipment of a facility, according to established criteria, such that the facility can withstand them without exceeding authorized limits.

 
Design Basis Accident
DBA A postulated accident leading to accident conditions for which a facility is designed in accordance with established design criteria and conservative methodology, and for which releases of radioactive material are kept within acceptable limits.



 
Design Basis External Event(s)
DBEE The external event(s) or combination(s) of external events considered in the design basis of all or any part of a facility.

 
Design Extension Conditions
DEC Postulated accident conditions that are not considered for design basis accidents, but that are considered in the design process of the facility in accordance with best estimate methodology, and for which releases of radioactive material are kept within acceptable limits. For nuclear power plants and research reactors, design extension conditions comprise conditions in events without significant fuel degradation and conditions in events with melting of the reactor core.
 
Deterministic safety analysis
Analysis using, for key parameters, single numerical values (taken to have a probability of 1), leading to a single value for the result.
In the safety of nuclear installations, for example, this implies focusing on accident types, releases of radioactive material and consequences, without considering the probabilities of different event sequences.
Typically used with either 'best estimate' or 'conservative' values, based on expert judgement and knowledge of the phenomena being modelled.

 
Deuterium
An isotope of hydrogen with one proton and one neutron in the nucleus.

 
Direct Containment Heating
DCH In a core melt accident, four mechanisms may cause a rapid increase of pressure and temperature in the reactor containment following a high-pressure melt ejection: (i) blowdown of the reactor coolant system; (ii) efficient debris-to-gas heat transfer; (iii) exothermic metal/steam and metal/oxigen reactions, (iv) hydrogen combustion. These processes, which lead to increased loads on the containment building, are collectively referred to as Direct Containment Heating. The pressure load imposed on the reactor containment building as a consequence of DCH may potentially lead to early failure of the containment.
 
Direct Current
DC In direct current, the flow of electric charge is only in one direction.
 

E

Emergency Control Coolant System
ECCS Reactor system components (pumps, valves, heat exchangers, tanks, and piping) that are specifically designed to remove residual heat from the reactor fuel rods in the event of a failure of the normal core cooling system (reactor coolant system).

 
Emergency Operating Procedure
EOP The strategies developed in the accident management programme should be converted to procedures for the preventive domain (EOPs) and guidelines for the mitigatory domain (SAMGs). Basically, procedures contain a set of actions to prevent the escalation of an event into a severe accident.

EOPs are plant procedures that direct operators' actions necessary to prevent the consequences of transients and accidents that have caused plant parameters to exceed reactor protection system set points or engineered safety feature set points, or other established limits.

In case smaller incidents occur, Nuclear Power Plants have instructions for the operators on what to do during these scenarios. Those are called Alarm Respose Procedures (ARP). See ARP definition.

 
Emergency Procedure Guideline
EPG Other name to be used for Emergency Operating Procedures. See EOP definition.

 
Emergency Response Organisation
ERO The NPP Organisation aimed at the performance of actions to mitigate the impact of an emergency on human health and safety, quality life, property or the environment.
 
Emergency Response Team
ERT The ERT is responsible for assessing the off-site consequences of an event and recommending off-site actions.
 
European Pressurized Reactor
EPR Evolutionary reactor manufactured by AREVA. See more information.

 
Extended Loss of Alternating Power
ELAP

 
Extensive Damage Mitigation Guidelines
EDMG EDMGs are guidelines adopted by some Member States that provide strategies to maintain or restore core cooling, containment, and SFP cooling capabilities under the circumstances associated with the loss of large areas of the plant due to explosions or fire.
 

F

Fission Product
FP A radionuclide produced by nuclear fission. Used in contexts where the radiation emitted by the radionuclide is the potential hazard.

 
Fission Product Barrier
FPB Nuclear Power Plants are designed with successive physical barriers against the release of fission products:
- Fuel matrix;
- Fuel cladding;
- Primary system pressure boundary;
- Containment.

See also Barrier definition.
 
Fuel rod
A rod of nuclear fuel, its cladding and any associated components necessary to form a structural entity.

 

H

High Pressure Injection System
HPIS The purposes of the High Pressure Coolant Injection (HPCI) System are to maintain adequate reactor vessel water inventory, for core cooling, on small break LOCAs, depressurize the reactor vessel to allow low pressure Emergency Core Cooling Systems (ECCSs) to inject on intermediate break LOCAs, and to backup the Reactor Core Isolation Cooling (RCIC) System under reactor vessel isolation conditions.

The HPIS system actuates automatically on low pressurizer pressure, high containment pressure, or when steam line pressure or flow anomalies are detected. Therefore, in addition to a LOCA, other events will lead to HPIS actuation. Some examples of such events are Steam Generator Tube Ruptures, RCS overcooling events resulting from steam line breaks (e.g.: stuck open main steam safety valves), or RCS depressurization events (e.g.: stuck open pressurizer spray valves).
 

I

Individual Plant Examination
IPE A risk analysis that considers the unique aspects of a particular nuclear power plant, identifying the specific vulnerabilities to severe accident of that plant.
 
Individual Plant Examination of External Events
IPEEE While the "individual plant examination" takes into account events that could challenge the design from things that could go awry internally (in the sense that equipment might fail because components do not work as expected), the "individual plant examination of external events" considers challenges such as earthquakes, internal fires, and high winds.
 
Interfacing Loss-of-Coolant Accident
ISLOCA An Interfacing Systems Loss-Of-Coolant Accident (ISLOCA) is a breach in a system that interfaces with the reactor coolant system (RCS) and could cause a loss of coolant accident, if the breach is not isolated from the RCS. Such a breach could be caused if valves fail to isolate the RCS from an interfacing system not designed for the high RCS pressures. When portions of an interfacing system are located outside the containment, particular concern arises because an unisolated system breach outside containment can result in a release of radionuclides that bypasses the containment.

 
Initiating event
IE An identified event that leads to anticipated operational occurrences or accident conditions.
This term (often shortened to initiator) is used in relation to event reporting and analysis, i.e. when such events have occurred. For the consideration of hypothetical events considered at the design stage, the term postulated initiating event is used.

Postulated initiating event (PIE). A postulated event identified in design as capable of leading to anticipated operational occurrences or accident conditions. The primary causes of postulated initiating events may be credible equipment failures and operator errors (both within and external to the facility), human induced events or natural events.
 
Isolation Condenser System
ICS The isolation condenser system in a boiling-water reactor (BWR) is a safety system that removes decay heat after reactor isolation during power operations. By removing decay heat, the system also limits reactor pressure increase to below the safety relief valve (SRV) setpoint and prevents SRV actuation following a scram. Upon actuation, the ICS provides water inventory that was held in the system piping to the reactor vessel. The ICS also provides decay heat removal (DHR) from the reactor core via the reactor coolant system (RCS) by natural circulation for an extended period without needing operator action.
 

K

Kerntechnischer Ausschuss
KTA The Nuclear Safety Standards Commission (KTA) has the task to issue nuclear safety standards for topics in the area of nuclear technology where a consensus between experts of the manufacturers and the operators of nuclear power plants, of authorized experts and state officials is apparent and to support their application.
 

L

Light Water Reactor
LWR A term used to describe reactors using ordinary water as coolant, including boiling water reactors (BWRs) and pressurized water reactors (PWRs), the most common types used worldwide.
 
Loss Of Coolant Accident
LOCA Those postulated accidents that result in a loss of reactor coolant at a rate in excess of the capability of the reactor makeup system from breaks in the reactor coolant pressure boundary, up to and including a break equivalent in size to the double-ended rupture of the largest pipe of the reactor coolant system.

 
Loss of normal access to the Ultimate Heat Sink
LUHS

 

M

Main Steam Isolation Valve
MSIV
 
Moderator
A material that is used to decrease the speed of the fast neutron produced during a fission even and thus sustaining a chain reaction.

 
Molten Core Concrete Interaction
MCCI The interaction of molten core material with the concrete during a core melt accident, which causes concrete ablation.
 

N

Normal operation
NO Operation within specified operational limits and conditions. For a nuclear power plant, this includes startup, power operation, shutdown, maintenance, testing and refuelling.
 
Nuclear Power Plant
NPP A nuclear power plant is a thermal power station in which the heat source is one or more nuclear reactors. As in a conventional thermal power station the heat is used to generate steam which drives a steam turbine connected to a generator which produces electricity.

 
Nuclear Power Technology Development Section
NPTDS With advice from representatives of Member States, the IAEA's Nuclear Power Technology Development Section fosters information exchange and collaborative research and development in the area of advanced nuclear reactor technologies needed to meet, in a sustainable manner, the increasing energy demands of the 21st century.

 
Nuclear Regulatory Commission
NRC The U.S. Nuclear Regulatory Commission (NRC) was created as an independent agency by Congress in 1974 to ensure the safe use of radioactive materials for beneficial civilian purposes while protecting people and the environment. The NRC regulates commercial nuclear power plants and other uses of nuclear materials, such as in nuclear medicine, through licensing, inspection and enforcement of its requirements.

 
Nuclear Steam Supply System
NSSS The reactor and the reactor coolant pumps (and steam generators for a pressurized water reactor) and associated piping in a nuclear power plant used to generate the steam needed to drive the turbine generator unit.
 

O

OECD Nuclear Energy Agency
OECD/NEA The Nuclear Energy Agency (NEA) is an intergovernmental agency that facilitates co-operation among countries with advanced nuclear technology infrastructures to seek excellence in nuclear safety, technology, science, environment and law.
The NEA, which is under the framework of the Organisation for Economic Co-operation and Development, is headquartered in Paris, France.
In order to achieve this, the NEA works as a forum for sharing information and experience and promoting international co-operation; a centre of excellence which helps member countries to pool and maintain their technical expertise and a vehicle for facilitating policy analyses and developing consensus based on its technical work.
The NEA's current membership consists of 33 countries in Europe, the Americas and the Asia-Pacific region.
 
Operational states
States defined under normal operation and anticipated operational occurrences. Some states and organizations use the term operating conditions (in contrast to accident conditions) for this concept.
 
Operational Support for Severe Accidents
OSSA Acronym for SAMG in France. See SAMG.
 
Operator
Any person or organization applying for authorization or authorized and/or responsible for safety when undertaking activities or in relation to any nuclear facilities or sources of ionizing radiation.
Operator includes, inter alia, private individuals, governmental bodies, consignors or carriers, licensees, hospitals, self-employed persons.

 
Outage
The shutdown of a generating unit, transmission line, or other facility for inspection, maintenance, or refueling, which is scheduled well in advance (even if the schedule changes). Scheduled outages do not include forced outages and could be deferred if there were a strong commercial reason to do so.

 
Owner
The company which owns a nuclear power plant majority.

 

P

Permanent shutdown
The cessation of operation of a facility with no intention to recommence operation in the future.

 
Pellet, Fuel
A thimble-sized ceramic cylinder, consisting of uranium (typically uranium oxide, UO2) or other fissile material (e.g. plutonium, etc.), to fuel a nuclear reactor. Modern reactor cores in pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) may contain up to 10 million pellets, stacked in the fuel rods that form fuel assemblies.

 
Plant equipment
Item important to safety. An item that is part of a safety group and/or whose malfunction or failure could lead to radiation exposure of the site personnel or members of the public.
Items important to safety include:
• Those structures, systems and components whose malfunction or failure could lead to undue radiation exposure of site personnel or members of the public;
• Those structures, systems and components that prevent anticipated operational occurrences from leading to accident conditions;
• Those features that are provided to mitigate the consequences of malfunction or failures of structures, systems and components.


Protection system. System that monitors the operation of a reactor and which, on sensing an abnormal condition, automatically initiates actions to prevent and unsafe condition.
This use of the term protection refers to protection of the plant. The system in this case encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device input terminals.

Safety actuation system. The collection of equipment required to accomplish the necessary safety actions when initiated by the protection system.

Safety related item. An item important to safety that is not part of a safety system.

Safety related system. A system important to safety that is not part of a safety system.
A safety related instrumentation and control system, for example, is an instrumentation and control system that is important to safety but which is not part of a safety system.

Safety system. A system important to safety, provided to ensure the safe shutdown of the reactor or the residual heat removal from the core, or to limit the consequences of anticipated operational occurrences and design basis accidents.
Safety systems consist of the protection system, the safety actuation and the safety system support features. Components of safety systems may be provided solely to perform safety functions, or may perform safety functions in some plant operational states and non-safety functions in other operational states.

Safety system support features. The collection of equipment that provides services such as cooling, lubrication and energy supply required by the protection system and the safety actuation systems.
 
Plant Damage Descriptors
PDC

 
Power Operated Relief Valves
PORV

 
Pressurized Heavy Water Reactors
PHWR Pressurized Heavy Water Reactors commonly use unenriched natural uranium as its fuel and heavy water (deuterium oxide D2O) as its coolant and moderator. See also CANDU definition.

 
Pressurised Water Reactor
PWR In a Pressurised Water Reactors, the reactor core heats water, which does not boil. This hot water then exchanges heat with a lower pressure water system, which turns to steam and drives the turbine.

The VVER (Water-Water Energetic Reactor), of Russian origin, is similar to PWRs but, among other features, the steam generator is horizontal.

 
Primary system
A term used to describe the reactor coolant system.

 
Probabilistic analysis
Often taken to be synonymous with stochastic analysis. Strictly, however, stochastic conveys directly the idea of randomness (or at least apparent randomness), whereas probabilistic is directly related to probabilities, and hence only indirectly concerned with randomness.
Therefore, a natural event or process might more correctly be described as stochastic (as in stochastic effect), whereas probabilistic events or processes and their consequences (such an analysis would, strictly, only be stochastic if the analytical method itself included an element of randomness, e.g. Monte Carlo analysis).
 
Probabilistic Safety Assessment
PSA A comprehensive, structured approach to identifying failure scenarios, constituting a conceptual and mathematical tool for deriving numerical estimates of risk.

Three levels of probabilistic safety assessment are generally recognized.
Level 1 comprises the assessment of plant failures leading to determination of the frequency of core damage.
Level 2 includes the assessment of containment response, leading, together with Level 1 results, to the determination of frequencies of failure of the containment and release to the environment of a given percentage of the reactor core's inventory of radionuclides.
Level 3 includes the assessment of off-site consequences, leading, together with the results of Level 2 analysis, to estimates of public risks.


 
Prototype reactor
Reactor in which much of the scaling up required for a commercial station in terms of both overall size and individual components has been incorporated.

 

Q

Quality Assurance
QA 1. The function of a management system that provides confidence that specified requirements will be fulfilled.

2. A systematic programme of controls and inspections applied by any organisation or body involved in the transport of radioactive material which is aimed at providing adequate confidence that the standard or safety prescribed in the Transport Regulation is achieved in practice.

3. All those planned and systematic actions necessary to provide confidence that a structure, system or component will perform satisfactorily in service.
 

R

RCC-M Design and Construction Rules
RCC-M AFCEN's RCC-M code concerns the mechanical components designed and manufactured for pressurised water reactors (PWR). It applies to pressure equipment in nuclear islands in safety classes 1, 2 and 3, and certain non-pressure components, such as vessel internals, supporting structures for safety class components, storage tanks and containment penetrations.
The design, fabrication and inspection rules defined in RCC-M leverage the results of the research and development work pioneered in France, Europe and worldwide, and which have been successfully used by industry to design and build PWR nuclear islands.
 
Reactor Coolant Pump
RCP The purposes of the RCP is: (i) To provide the forced circulation of reactor coolant for the removal of core heat, (ii) Improve departure from nucleate boiling ratio during loss of all reactor coolant pump motor power, (iii) To provide energy to heat up the RCS from ambient temperature to greater than the minimum temperature for criticality prior to reactor startup.
 
Reactor Coolant System
RCS The system used to remove energy from the reactor core and transfer that energy either directly or indirectly to the steam turbine.

 
Reactor core
The central portion of a nuclear reactor, which contains the fuel assemblies, moderator, neutron poisons, control rods, and support structures. The reactor core is where fission takes place.

 
Reactor Pressure Vessel
RPV A strong-walled container housing the core of most types of power reactors. It usually also contains the moderator, neutron reflector, thermal shield, and control rods.

 
Refueling Water Storage Tank
RWST Water source in Nuclear Power Plants that provides water to the Emergency Core Cooling System.
 
Residual Heat Removal System
RHRS The purposes of the residual heat removal system are as follows: (i) Removes decay heat from the core and reduces the temperature of the RCS during the second phase of plant cooldown, (ii) Serves as the low pressure injection portion of the Emergency Core Cooling System (ECCS), following a loss of coolant accident, and (iii) Transfers refueling water between the refueling water storage tank and the refueling cavity before and after refueling.
The RHRS transfers heat from the reactor coolant system to the component cooling water system. During shut down plant operations, the RHR system is used to remove the decay heat from the core and reduce the temperature of the reactor coolant to the cold shutdown temperature.
 

S

Safeguards
The material control and accounting program which controls the enriched nuclear material. As used by the International Atomic Energy Agency, this term also means verifying that the peaceful use commitments made in binding nonproliferation agreements, both bilateral and multilateral, are honored.
 
Safety Relief Valve
SRV
 
Safety Assessment Section
SAS The Safety Assessment Section works with international experts to develop new and revise existing safety standards, with the goal to assist the IAEA Member States in achieving and maintaining a high level of safety. To complement the IAEA Safety Standards, the Section also provides technical safety review services to Member States.

 
Severe Accident
Accident more severe than a design basis accident and involving significant core degradation.

 
Severe Accident Management
SAM The taking of a set of actions during the evolution of a severe accident, it includes measures to:
- Terminate the progress of core damage once it has started;
- Maintain the integrity of the containment as long as possible;
- Minimize releases of radioactive material.

 
Severe Accident Management Guideline
SAMG If an accident occurs at a nuclear power plant, to restore safety, two types of accident management guidance document are typically used: emergency operating procedures (EOPs) for preventing fuel rod degradation, and severe accident management guidelines (SAMGs) for mitigating significant fuel rod degradation when a severe accident is imminent.
The development of SAMGs is an essential part of the severe accident management programme.

 
Severe Accident Management Guideline Development Toolkit
SAMG-D The SAMG-D describes the elements necessary to develop a full package of Severe Accident Management Guidelines, which serve to achieve the main goals of severe accident management at a Nuclear Power Plant.

 
Shutdown
The cessation of operation of a facility.

 
Single failure
A failure which results in the loss of capability of a single system or component to perform its intended safety function(s) and any consequential failure(s) that result from it. A system is designed against a single failure if it is capable of performing its task in the presence of any single failure.
 
Spent Fuel
SF 1. Nuclear fuel removed from a reactor following irradiation that is no longer usable in its present form because of depletion of fissile material, poison buildup or radiation damage.
2. Nuclear fuel that has been irradiated in and permanently removed from a reactor core.
 
Spent Fuel Pool
SFP An underwater storage and cooling facility for spent (depleted) fuel assemblies that have been removed from a reactor.
 
Structures, Systems and Components
SSC A general term encompassing all of the elements (items) of a facility or activity which contribute to protection and safety, except human factors.

- The structures are the passive elements: buildings, vessels, shielding, etc.
- A system comprises several components, assembled in such a way as to perform a specific (active) function.
- A component is a discrete element of a system. Examples of components are wires, transistors, integrated circuits, motors, relays, solenoids, pipes, fittings, pumps, tanks and valves.
 

T

Technical Support Centre
TSC
 
Technical Support Guidelines
TSG
 

U

Unit
In terms of nuclear energy a unit is a single reactor at a multi-reactor nuclear power plant.

 
Uranium
Element No.92 in the periodic table. Found in natural ores and certain isotope can be used as nuclear fuel.

 

V

VVER
Vodo-Vodyanoi Energetichesky Reactor; Water-Water Energetic Reactor.